THE SECOND SAUDI INTERNATIONAL CONFERENCE ON NUCLEAR POWER ENGINEERING (SCOPE-2)

This is an abstract of the paper presenting a thermal-mechanical analysis based on ANSYS solutions, which was performed for two proposed designs of the component separating the cold inlet and heated outlet flows in the CSCWR concept: a single-wall component (monolithic wall) and a double-wall component incorporating a porous structure (composite wall).

This study explores the potential role of nuclear energy in supporting global efforts to reach net-zero carbon emissions, using Bangladesh’s Rooppur Nuclear Power Plant (RNPP) as a case example. As international focus intensifies on sustainable energy solutions, nuclear power is increasingly recognized for its low greenhouse gas emissions and stable output compared to intermittent renewable sources. The paper begins by outlining the broader global context, emphasizing nuclear energy's advantages, including minimal lifecycle emissions and strong safety records. The RNPP, located in Pabna, Bangladesh, is the country’s first nuclear energy facility, comprising two VVER-1200 reactors with a combined capacity of 2.4 gigawatts. It is expected to play a pivotal role in reducing reliance on fossil fuels and improving national energy security. The analysis compares projected emissions from RNPP to those of traditional fossil fuel-based plants such as coal, natural gas, HFO, and diesel. The findings reveal that RNPP’s CO₂ emissions would be drastically lower by about 98% and emissions of SO₂ and NOₓ would be nearly eliminated. Furthermore, the study uses international data to demonstrate that nations with a higher nuclear power share tend to have better air quality (lower AQI) and stronger environmental sustainability scores (higher EPI). It also highlights the spatial efficiency of nuclear facilities, showing that RNPP will require significantly less land per unit of energy compared to coal-based plants. In conclusion, the paper argues that nuclear energy, alongside renewables, can significantly contribute to cleaner air and reduced carbon output. The RNPP is presented as a model for how nuclear power can aid developing nations in meeting their climate goals while enhancing energy reliability. This case study reinforces the broader message that nuclear energy should be considered a key technology in the global transition to low-carbon energy systems.

Rapid industrialization has increased the demand for electricity in Bangladesh. In addition, it has a population of over 180 million, with minimal energy resources. Thus, Bangladesh is considering nuclear power to meet the growing energy demands. It has already constructed its first nuclear power plant with 2400 MW capacity in Rooppur, Pabna, and the government is planning more nuclear power plants in the future.

It is crucial to acknowledge public perception and social awareness regarding nuclear power and its safety, technology advancements, energy justice, awareness, etc. Several countries, such as the US, China, and India, have conducted a thorough study on the community acceptance of nuclear power, energy justice, and equity. There is a lack of proper highlights regarding Bangladesh’s first nuclear power plant.

This study aims to explore people’s perceptions and acceptance of nuclear power by considering several socioeconomic and technical factors through a series of structured and semi-structured interviews. The outcomes of this investigation will contribute to shaping a country’s energy policy and justice and the stakeholder’s role of engagement in nuclear power.

I. Introduction

Medical cyclotrons (MCs) have undergone significant technological evolution since their inception. Initially large and fixed-energy systems, they have evolved into compact, high-efficiency devices tailored for hospital-based applications. In Saudi Arabia, the integration of MCs into the national healthcare infrastructure reflects this broader trend and underscores the need for sustained optimization. Optimization of the cyclotron lifecycle, from design and shielding to isotope production, maintenance, and decommissioning, is essential for safe, cost-effective, and regulation-compliant operations [3][4]. This review article aims to examine the complete lifecycle of MCs deployed in Saudi Arabia, highlighting key practices across installation, operation, optimization, and decommissioning.

II. The Life Cycle of Medical Cyclotrons in Saudi Arabia
MC facilities in Saudi Arabia follow a structured workflow comprising
II.A. Installation Phase
Involving infrastructure preparation, shielding design, regulatory approval, and commissioning. According to IAEA guidance, planning for installation must begin with clear definitions of the intended radionuclide outputs, anticipated workload, and regulatory constraints.
II.B. Maintenance and Quality Control Phase
MC facilities in Saudi Arabia must adhere to a robust maintenance and quality control (QC) regimen to ensure safety, reproducibility, and GMP compliance.
II.C. Decommissioning Phase
Decommissioning is increasingly recognized as a critical lifecycle phase requiring early planning. Exposure to secondary neutrons during years of operation leads to the activation of cyclotron vault concrete and internal metal components.

III. Clinical and Industrial Impact
Cyclotrons are vital for theranostics, enabling targeted radioligand therapy and imaging. The expansion of tracer applications in neurology and cardiology underscores the broader relevance of MCs beyond oncology.

IV. Conclusions
Medical cyclotrons are at the heart of nuclear medicine innovation in Saudi Arabia. From installation to decommissioning, their lifecycle requires multidisciplinary collaboration and regulatory alignment. Scaling up isotope production and tracer innovation is essential to support future-ready healthcare systems.

This work presents a preliminary neutronic analysis of the SMART (System-integrated Modular Advanced Reactor), a small modular integral PWR developed by KAERI. The study, conducted at the Nuclear Engineering Department of the Federal University of Minas Gerais (DEN-UFMG), aims to support the future integration of thermal-hydraulic and neutronic simulations for accident-tolerant fuel (ATF) analysis in SMRs. In this initial phase, the macroscopic cross sections of the SMART reactor core were generated using the deterministic lattice code WIMSD-5B and then provided as input to the PARCS core simulator to evaluate reactor behavior under steady-state conditions.
The SMART reactor design includes 57 fuel assemblies, divided into four fuel types (A, B, C, D), each modeled as a homogenized unit cell composed of concentric annular regions. Fuel rods consist of UO₂ enriched to 4.95 w/o of U-235, gadolinia-bearing rods (Gd₂O₃–UO₂), and Al₂O₃–B₄C control rods with B-10 enriched to 30 w/o. Based on KAERI reference documents, reactor geometry and material composition were processed for two-group cross section generation. These were used to feed the PARCS code, along with kinetic parameters, to simulate the reactor core.
The results yielded consistent macroscopic cross section data and an effective multiplication factor of keff = 1.158462, indicating supercriticality and homogeneous power distribution, in line with SMART design references. Although preliminary, these results validate the modeling approach and support future coupling with RELAP5 for transient safety analysis and ATF performance assessment.

This research presents the first application of Sparse Identification of Nonlinear Dynamics (SINDy) to nuclear reactor model identification, specifically targeting Pressurized Water Reactor systems. The identified models provide a foundation for advanced control systems, predictive diagnostics, and operational optimization while maintaining the interpretability required for nuclear safety applications. The approach addresses traditional first-principles modeling limitations by discovering governing equations directly from measurements without requiring simplifications that may compromise predictive accuracy during complex operational transients.

A paper that presents the learning process of using the HCP, parameterized through benchmarking against several existing pebble bed reactor designs such as HTR-10 and HTR-PM. Subsequently, an analysis was conducted on several parameters, including excess reactivity, radionuclide content, and pebble fuel consumption, for each variation of reactor geometry design in order to determine the most optimal design.

The ongoing global energy crisis, driven by increasing demand and growing environmental concerns, has intensified the need for sustainable and integrated solutions particularly in the field of water desalination. Desalination technologies are known for their high energy requirements and reliance on fossil fuels, resulting in elevated operational costs and significant environmental impact. In contrast, membrane distillation (MD) has gained recognition as a viable alternative due to its compatibility with low-grade heat sources. Nuclear power plants, although primarily engineered for electricity generation, discharge a substantial amount of thermal energy, especially through steam exhausted from the low-pressure turbine stage. This waste heat is typically underutilized, representing a missed opportunity for energy recovery. In this study, the feasibility of integrating this residual thermal energy with a spiral-wound membrane distillation system is investigated. The main analysis parameters are varying steam extraction pressures on key performance metrics of the MD process, including permeate flux and gained output ratio (GOR). This integrated approach not only addresses the need for more efficient solutions but also demonstrates the potential for improved energy utilization in nuclear facilities.

Study involving the modeling of various SGTR accident scenarios (including MSGTR) using the MELCOR thermal-hydraulic code. The analysis focuses on quantifying the potential radiological release and assessing the risk of core degradation. The results aim to enhance the understanding of accident progression and core damage risk associated with tube rupture scenarios.

With the growing global emphasis on low-carbon and energy-secure systems, nuclear energy is becoming a stable electricity source and a reliable supplier of industrial-grade heat. Nuclear energy could enable the hydrogen economy through its generation of hydrogen by thermochemical or electrochemical methods [1]. Integrating hydrogen production facilities with nuclear power plants presents distinct safety issues that must be comprehensively evaluated for the combined operation to be viable [2] [3]. This study uses qualitative and quantitative approaches to systematically understand the potential risks throughout heat extraction, hydrogen generation, and operations on a shared site. The results of this work are intended to inform future licensing efforts by providing necessary risk insights to regulators, designers, and stakeholders looking to implement nuclear-assisted hydrogen technology safely and responsibly.

The increasing need for efficient and reliable radiation monitoring in nuclear facilities has driven the development of advanced early warning systems. Conventional Radiation Early Warning Systems (REWS) often rely on manual validation, which can delay critical responses during emergencies. Recent nuclear incidents, such as Fukushima, highlight the importance of real-time, automated monitoring and intelligent evacuation planning to minimize radiation exposure and ensure public safety.

This work introduces GAMOWS, an Artificial Intelligence of Things (AIoT)-based module designed to provide real-time radiation monitoring and intelligent evacuation route recommendations for nuclear facilities. The GAMOWS module integrates several key components and algorithms to achieve these goals, including IoT-based sensors for data acquisition, AI-driven predictive learning for risk assessment, and a modified Dijkstra algorithm for evacuation path planning. Field test simulation results show that the most effective exit route ends at the nearest evacuation path, with a total accumulated dose of 1.355 × 10⁻³ µSv, a travel distance of 19.49 meters, and a required time of 8.77 seconds while running. When the radiation source is hypothesized at the center (reactor core), simulations indicate that the optimal evacuation route from the front door to the nearest exit accumulates a dose of 4.853 × 10⁻³ μSv over a distance of 66.38 meters in 29.8 seconds.

Overall, GAMOWS demonstrates the potential of AIoT-based systems for enhancing radiation monitoring and emergency response in nuclear facilities, where the integration of Artificial Neural Networks (ANN) and intelligent pathfinding enables real-time risk assessment and optimized evacuation, supporting safer and more resilient nuclear operations.

Dear representatives of SCOPE-2,

I am sending the extended abstract files in .pdf and .docx format as an attachment.
1) Brzezinski_Mikolaj_KINGS_SCOPE-2_Extended_Abstract.docx
2) Brzezinski_Mikolaj_KINGS_SCOPE-2_Extended_Abstract.pdf

Moreover, I would like to kindly highlight my desire to participate in the "Travel Grant for Students program", covering tickets, accommodation, and registration fee.

Sincerely,
Mikolaj Brzezinski

The research represents a groundbreaking application of artificial intelligence to nuclear fuel assembly optimization, specifically targeting the NuScale Small Modular Reactor design . The work addresses critical challenges in nuclear fuel cycle optimization by developing an automated, physics-informed reinforcement learning framework that significantly outperforms traditional optimization methods

The growing interest in compact and lightweight nuclear reactors, particularly Small Modular Reactors (SMRs), calls for innovative and simplified control strategies. This study investigates the viability of using rotating annular reflector drums as a primary control mechanism in SMRs an approach not applied in Pressurized Water Reactor (PWR) technology previously. The concept involves modulating neutron economy through varying reflector materials and geometries instead of traditional control rods. A detailed 3D reactor model is being developed using advanced computational tools to perform high-fidelity neutron transport simulations. The study evaluates combinations of reflector materials, fuel types, and moderators to optimize reactivity control. Additionally, the dynamic behaviour of neutron generation density is analysed across various operational phases of drum rotation.
This control strategy not only simplifies reactor design but also inherently eliminates the risk of rod ejection accidents, offering a safer and potentially more robust control mechanism for next-generation SMRs.

[1] Malloy, J., Jacox, M. and Zubrin, R., 1992, July. Small Externally-fueled Heat Pipe Thermionic Reactor (SEHPTR) for dual mode applications. In 28th Joint Propulsion Conference and Exhibit (p. 3585).
[2] Gul, Anas, R. Khan, M. Azeem, I. Shahzad, and T. Stummer. "Benchmarking of Monte Carlo model of PWR (CNPP-II) core against theoretical and experimental results." Progress in Nuclear Energy 92 (2016): 164-174.
[3] El Yaakoubi, H., et al. "Neutronic modeling and calculation of the Nuclear Heating Reactor NHR-5 by the deterministic codes DRAGON5 & DONJON5." Progress in Nuclear Energy 142 (2021): 104000.

The Indonesian government aims to incorporate nuclear energy into its Net Zero Emission 2060 strategy, with nuclear power plant operations projected to begin by 2032. The Sodium Fast Reactor (SFR) system stands out across various aspects. However, using liquid sodium as a coolant in SFRs presents thermohydraulic safety challenges, particularly in Unprotected Loss of Flow (ULOF) scenarios. Previous studies have shown that ULOF can lead to centralized sloshing, increasing the risk of recriticality.

This study evaluates the Moving Particle Semi-implicit (MPS) method for simulating the centralized sloshing phenomenon in SFRs, using Maschek's (1992) experiment as a benchmark. The research incorporates modifications to the MPS method proposed by Kondo and Koshizuka (2020) to address known limitations of the classic MPS algorithm, such as pressure fluctuations and singularity issues in the weighting function. Initial surveys indicate that smaller particle sizes yield more accurate results. The 'WP Modification' within Kondo's MPS improvements significantly enhances accuracy by improving particle compactness and thrust. The study also includes a size parameter survey and a 2D beta and gamma survey for time efficiency, comparing the best beta and gamma formulations with classical methods based on accuracy, pressure analysis, velocity vectors, and non-dimensional flow characteristics. The optimal results are then compared with the Smoothed Particle Hydrodynamics (SPH) method. The findings suggest that the MPS method with Kondo's improvements effectively simulates centralized sloshing, leading to denser particle motion and more accurate pressure, with accuracy dependent on the proper selection of beta and gamma parameters.

Boiling curves, which relate heat flux density to surface superheating, are traditionally derived under stationary conditions and widely used for modeling heat exchange in power equipment. However, real-world scenarios often involve rapid increases in heat flux, such as in reactor power surges or superconducting systems. This study experimentally investigates the differences between boiling curves obtained under stationary and nonstationary heat release conditions in water at atmospheric pressure using a thin horizontal cylindrical heater. The results demonstrate a shift of the boiling curve toward higher dissipated heat fluxes with increasing heating rates. The paper quantifies the deviation between stationary and nonstationary curves and identifies conditions where stationary correlations remain applicable.

Boiling flows play a pivotal role in a wide array of industrial processes, particularly in energy systems such as nuclear reactors, where efficient thermal management is essential to ensure system performance and safety. These flows exhibit complex behaviour due to the interaction of multiple phases, and they are characterized by a variety of flow regimes that depend on the local phase distribution pattern. Among the most challenging of these regimes is slug flow boiling, where large vapour structures form near heated surfaces, significantly influencing interfacial dynamics and heat transfer mechanisms. In recent years, the development of computational fluid dynamics (CFD) has greatly advanced the ability to simulate and analyse complex multiphase flows. However, conventional two-fluid models often struggle to accurately predict flow behaviour in regions with varying interfacial scales, particularly where the transition from dispersed bubbly flow to slug or annular flow occurs. To address these challenges, the Generalised Multifluid Modelling Approach (GEMMA), has been developed and implemented within the OpenFOAM code. This methodology enables dynamic selection between interface-resolving and Eulerian-Eulerian approaches within each computational cell, depending on the scale of interfacial structures present, improving the prediction of multiscale phenomena This study presents a significant extension to the GEMMA model by incorporating enhanced capabilities for simulating subcooled boiling flows across a wide range of vapour volume fractions. Specifically, the focus is on accurately capturing the transition from low void fraction nucleate boiling to high void fraction slug flow near heated walls, with emphasis on accurately predicting heat and mass transfer in the transition from nucleate boiling to vapour slug flow.

Using the concept of Prompt Gamma Neutron Activation Analysis (PGNAA), we conducted monte carlo simulations (MCNPX) to determine whether measurement of formation porosity can be done using a neutron source and two gamma detectors for the deep borehole disposal (DBD) purpose by using abandoned oil wells. The gamma spectrums in three different simulation conditions then analyzed to measure the sensitivity of the tool. Three variations of condition as mentioned above are, applying different neutron sources, adding wax into the borehole inner layer, and applying boron-lining outside the detector. To validate this research, comparison between the simulation result using MCNPX and the lab result have been conducted. It is possible to measure formation porosity by utilizing two BGO (Bismuth Germanate Oxide) detectors and a neutron source. In conclusion, to gain the best tool's sensitivity, we suggest the use of AmBe (Americium Beryllium) as the neutron source rather than DT (Deutrium Tritium) neutron generator without boron-lining attached to the detector. From the gamma spectrums analysis, defining detector count as the summation of count in all energy ranges provides the most stable sensitivity rather than calculating certain peaks. Whereas wax presence does not significantly affect the sensitivity.

Dr. Fatai Liadi will present the work

An Automatic Soluble Boron Scenario Based on a Sliding Mode Observer During Daily Load Follow Operation in APR1400

The safe decontamination of abandoned radioisotope and radiopharmaceutical production facilities in Indonesia for reuse is a pressing need due to the increasing local demand and the country's current dependence on imports. This study presents a structured risk evaluation model for the decontamination process in such facilities, integrating qualitative risk assessment with the Analytic Hierarchy Process (AHP) as a deterministic machine learning approach. The model prioritizes risk control strategies based on five critical criteria: effectiveness, cost, implementation time, regulatory compliance, and environmental impact. AHP was employed to calculate the relative weights of each criterion and evaluate four alternative risk control measures through pairwise comparisons and eigenvector calculations. The results identified radiation shielding (Alternative A1) as the most effective control measure, scoring 0.3925, significantly outperforming the other alternatives. This prioritization framework reflects a machine learning-aligned decision-making process, enabling data-structured reasoning for safety and regulatory adherence. While the current method is deterministic, its architecture supports the future integration of dynamic, data-driven machine learning models that are capable of real-time risk prediction and adaptive control strategy. Overcoming current limitations, such as static input assumptions and limited operational feedback, could pave the way for intelligent, automated decision-support systems that enhance radiation protection and operational efficiency in high-risk environments.

The revitalization of abandoned nuclear facilities requires a strategic approach that emphasizes safety, operational efficiency, and sustainability. Robotics technology has emerged as a transformative solution, offering remote operation capabilities that reduce human exposure to hazardous environments while increasing precision and effectiveness in revitalization tasks. International examples demonstrate the potential of repurposing nuclear sites for research and industrial use through robotics-enabled interventions.
Radiation mapping is a foundational task during revitalization. Traditional manual methods are labor-intensive, hazardous, and offer limited spatial resolution. Robotic systems equipped with gamma sensors and advanced path-planning algorithms enable efficient, high-resolution radiation mapping in complex environments. Planning strategies that integrate coverage path planning and real-time data acquisition significantly enhance mapping performance while maintaining safety.
In the cleaning and dismantling phase, robotics facilitates safe handling of contaminated materials. Technologies such as remote manipulators, autonomous vehicles, and hybrid robots extend capabilities to navigate confined, unpredictable environments. These systems mitigate risks posed by radiation, structural degradation, and chemical hazards, while modular and adaptable configurations improve task versatility.
Despite these advancements, technical limitations remain. Challenges include radiation-induced degradation, navigation in unstructured environments, and integration with digital infrastructure. Addressing these requires robust design, radiation-hardened components, and intelligent control frameworks. Future strategies should incorporate digital twin environments, AI-assisted planning, and human-in-the-loop systems to improve adaptability and coordination.
This review presents a roadmap for robotics planning in nuclear facility revitalization, synthesizing best practices and identifying innovation pathways. Implementing such strategies can convert high-risk legacy sites into productive, safe, and sustainable assets, contributing to environmental restoration and socio-technical resilience in the nuclear sector.

The increasing number of disused sealed radioactive sources (DSRS), particularly from cobalt-60 (⁶⁰Co) teletherapy units, poses a significant challenge for radioactive waste management in Indonesia. In accordance with national regulations, these sources must be securely transported and stored by the authorised institution. However, the lack of cost-effective and certified transport containers has led to suboptimal interim storage practices by hospitals, such as using teletherapy machine heads as makeshift containment units.
To address this issue, a specialised transport container, referred to as the Gama Container, has been developed with integrated radiation shielding. This study assesses the shielding performance of the container through simulation using MicroShield and the Particle and Heavy Ion Transport code System (PHITS). The shielding structure, designed in a cylindrical geometry with lead and concrete layers, was evaluated for a 6600 Ci ⁶⁰Co source. Dose rate calculations were performed across various shielding thicknesses and distances.
MicroShield simulations demonstrated that a 30 cm concrete layer reduced the external dose rate from 17.98 µSv/h to 0.01 µSv/h at 1 metre. PHITS simulation results supported these findings, indicating spatial dose distributions ranging from 0.008 to 0.012 µSv/h outside the container. These values fall well below national exposure limits for both occupational and public safety.
The study confirms that the Gama Container's design meets regulatory safety standards, ensuring secure transportation of high-activity DSRS. The integration of deterministic and Monte Carlo-based analyses provides a robust evaluation framework, supporting its practical deployment in the field. This approach contributes to enhancing the safety, reliability, and compliance of radioactive waste transport infrastructure in emerging nuclear technology environments.

Flow induced vibration (FIV) is a known problem
in nuclear reactors which exist due to the interaction
between the turbulent flow of the coolants and the fuel
rod bundles generating vibrations which are
undesirable. These vibrations can be a major cause of
fatigue failures, stress corrosion cracking and fretting
wear of materials, which lead to stand-still costs. To
numerically model the interaction between the fluid
and the structure, the system should couple the solvers
for both the phases. Coupling is achieved by mapping
the forces and displacements at the interface of fluid
and structure domains, also known as the partitioned
coupling.

The Sustainable Nuclear Energy (SNE) research team is designing a Molten Salt Reactor (MSR) for repowering coal power plants. MSR is an Generation IV reactor, uses liquid fuel composed of fluoride or chloride salt mixtures, enabling high-temperature operation at low pressure [1]. Optimal flow distribution is essential to ensure uniform power and efficient heat transfer across all fuel channels [2]. Several previous studies [2], [3], and [4] have calculated the flow rate ratio of each channel relative to the average flow rate. In this context, the focus of the present study is placed on assessing the performance of the reactor’s hydraulic structure in achieving uniform fluid flow distribution. Therefore, this study aims to design an optimized MSR–Repowering system with improved uniformity in fluid flow distribution.

The flow uniformity characteristic is determined by the mass flow rate entering each individual fuel pin (mi), the average mass flow rate per fuel pin (ma), and the total number of fuel pins (n).

This study successfully reduced flow maldistribution, with Sreactor decreasing from 63.3% to 54.96% and Sring from 5.0% to 4.89%. Further research will install shrouds to significantly reduce maldistribution.

The application of advanced fluids with improved thermal properties is a potential approach to enhance heat transfer efficiency in nuclear reactors, resulting in an elevation in the safety margin. This study investigates the impact of a hybrid TiO₂-Cu/water nanofluid on the thermodynamic behavior of coolant flow within a triangular array of a PWR subchannel. A single-phase flow of a homogeneous mixture of species was assumed. The analysis considers Reynolds numbers ranging from 20,000 to 80,000, in increments of 20,000, and evaluates three nanoparticle volume ratios: 1% TiO₂–3% Cu, 2% TiO₂–2% Cu, and 3% TiO₂–1% Cu. Key performance parameters analyzed include variations in the Nusselt number, convective heat transfer coefficient, temperature distribution, enthalpy gain, shear stress, and pressure drop. This study exhibited that the addition of nanoparticles enhances the overall convective heat transfer coefficient by an average of 5.2% compared to the base fluid (water). A higher volume fraction of TiO₂ improves heat transfer performance, although the rate of improvement relative to the water decreases as the Reynolds number increases. Among all the combinations, the 3% TiO₂–1% Cu ratio demonstrates the most efficient heat transfer based on temperature profile and Nusselt number trends. However, it also incurs a notable pressure drop. In contrast, the 1% TiO₂–3% Cu combination yields a reduction of 4.31% on average in pressure drop compared to the water. The flow dynamics and heat dissipation characteristics of the hybrid nanofluid were thoroughly examined using CFD simulations in ANSYS Fluent.

The Indonesian government aims to incorporate nuclear energy into its Net Zero Emission 2060 strategy, with nuclear power plant operations projected to begin by 2032. The Sodium Fast Reactor (SFR) system stands out across various aspects. However, using liquid sodium as a coolant in SFRs presents thermohydraulic safety challenges, particularly in Unprotected Loss of Flow (ULOF) scenarios. Previous studies have shown that ULOF can lead to centralized sloshing, increasing the risk of recriticality.

This study evaluates the Moving Particle Semi-implicit (MPS) method for simulating the centralized sloshing phenomenon in SFRs, using Maschek's (1992) experiment as a benchmark. The research incorporates modifications to the MPS method proposed by Kondo and Koshizuka (2020) to address known limitations of the classic MPS algorithm, such as pressure fluctuations and singularity issues in the weighting function. Initial surveys indicate that smaller particle sizes yield more accurate results. The 'WP Modification' within Kondo's MPS improvements significantly enhances accuracy by improving particle compactness and thrust. The study also includes a size parameter survey and a 2D beta and gamma survey for time efficiency, comparing the best beta and gamma formulations with classical methods based on accuracy, pressure analysis, velocity vectors, and non-dimensional flow characteristics. The optimal results are then compared with the Smoothed Particle Hydrodynamics (SPH) method. The findings suggest that the MPS method with Kondo's improvements effectively simulates centralized sloshing, leading to denser particle motion and more accurate pressure, with accuracy dependent on the proper selection of beta and gamma parameters.

The separation of fine particles from aqueous suspensions is a major challenge in radwaste management, particularly during the treatment of sludges from aging facilities. These sludges often contain stable micron-sized solids that hinder dewatering. Retrieval operations can exacerbate colloid persistence, increasing the risk of radionuclide mobility and compromising facility safety. Conventional methods often fail to achieve the clarity or compaction required for effective processing. As a solution, high molecular weight polyacrylamide (PAM) flocculants are widely applied to enhance particle aggregation, accelerate settling, and reduce supernatant turbidity. Polymeric flocculation, typically used in wastewater treatment, has shown promise in improving radwaste dewatering.

This study evaluates a suite of PAM flocculants for enhancing the solid-liquid separation of calcite-based suspensions that simulate the behaviour of inorganic nuclear waste sludges. The goal is to understand flocculation performance under variable shear conditions and the corresponding sedimentation and flotation performance. This work will help support efficient legacy radwaste treatment.

This study demonstrates that PAM can significantly enhance the separation of calcite suspensions representative of radwaste sludges. Optimal sedimentation was achieved using A-PAM, producing stable flocs and effective turbidity reduction. However, these systems exhibited limited flotation. In contrast, C-PAM, while less effective for settling, delivered the highest flotation recovery, likely due to favorable interactions with the anionic collector under alkaline conditions. These findings support the use of tailored flocculant selection to balance sedimentation and flotation goals in radwaste dewatering applications.

Abstract: High integrity container (HIC) has become a sustainable solution for radioactive spent ion exchange resins (IERs) conditioning due to its stability for long-term storage and disposal and has been implemented in Wolseong Korean nuclear power plant (NPP) and some China’s NPPs. Despite the absence of NPP in Indonesia, radioactive IERs in Indonesia has been generated from nuclear research reactors, pool storage for spent fuels, and liquid radioactive waste treatment, emphasizing the need for sustainable conditioning particularly in using HIC. This study examines global research trends in High Integrity Container (HIC) technology for conditioning spent radioactive ion exchange resins (IERs). Using bibliometric analysis of publications and patents (1960-2024), we identify key developments and gaps in the field. The results highlight significant disparities in research contributions, with developed nations leading innovation while developing countries remain underrepresented. Our findings provide guidance for future research priorities, particularly in adapting HIC technology for non-power reactor applications. The study underscores the need for international collaboration to advance sustainable radioactive waste management solutions.

This study addresses the critical issue of nuclear fuel disposal by proposing a comprehensive multi-objective optimization framework that balances cost, safety, environmental impact, and disposal time. Unlike traditional approaches that primarily rely on heuristic algorithms—such as Genetic Algorithm (GA), Particle Swarm Optimization (PSO), Grey Wolf Optimization (GWO), Aquila Optimization Algorithm (AOA), and Simulated Annealing (SA)—this research integrates a non-linear exact solver, the Interior Point Optimizer (IPOPT), to benchmark and compare performance. Each algorithm is assessed using a unified model, with heuristic methods undergoing dedicated hyperparameter tuning to ensure fairness. IPOPT delivers the most optimal solution, excelling across all objectives due to its mathematical precision, although at the cost of higher computational time. Among heuristic algorithms, Simulated Annealing demonstrates the fastest performance with strong results, while GA and PSO perform identically, ranking second in effectiveness. AOA and GWO follow with viable but less optimal outputs. The study highlights the trade-off between computational speed and solution quality, underscoring the need for context-driven algorithm selection in real-world applications. By combining exact and heuristic methods, the research offers a robust and scalable strategy for sustainable nuclear waste management, providing valuable insights for engineers, decision-makers, and policy planners in the energy and environmental sectors.

The rising global demand for rare earth elements (REEs), especially in advanced nuclear energy systems, has increased interest in alternative and sustainable sources. Phosphogypsum (PG), a byproduct of phosphate fertilizer production, contains notable amounts of REEs such as lanthanum, uranium, thorium, cerium, neodymium, samarium, and dysprosium—elements critical to nuclear fuel applications. This study evaluates PG as a secondary source of these materials by focusing on its radiochemical and elemental composition. Samples from three PG sources, Wizów, Police, and Wiślinka (Poland), were analyzed to assess variations in radionuclide content and REE concentrations. High-purity germanium (HPGe) gamma and alpha-particle spectrometry were employed to quantify radionuclides from the 238U and 232Th decay series and 40K based on IAEA methodologies. Results revealed significant differences in activity levels between samples. The chemical composition was determined by ICP-MS and XRF, detecting REEs in all samples, with Police PG showing notably higher concentration. This research demonstrates the viability of PG as a REE-bearing material and potential input for nuclear-related applications by determining the precipitating methods of REEs. The study’s primary objective is to assess PG as a dual-purpose material, addressing environmental waste management and the sustainable supply of critical resources for the nuclear fuel cycle. The work reduces REE extraction's environmental impact and carbon footprint by identifying PG as an alternative to traditional mining. This approach aligns with circular economy principles in the nuclear sector and advances long-term sustainability goals, including CO₂ emission reduction.
Keywords:
Phosphogypsum, Rare Earth Elements, Nuclear Fuel, Radioactivity, Environmental Waste Management

The rising global demand for rare earth elements (REEs), especially in advanced nuclear energy systems, has increased interest in alternative and sustainable sources. Phosphogypsum (PG), a byproduct of phosphate fertilizer production, contains notable amounts of REEs such as lanthanum, uranium, thorium, cerium, neodymium, samarium, and dysprosium—elements critical to nuclear fuel applications. This study evaluates PG as a secondary source of these materials by focusing on its radiochemical and elemental composition. Samples from three PG sources, Wizów, Police, and Wiślinka (Poland), were analyzed to assess variations in radionuclide content and REE concentrations. High-purity germanium (HPGe) gamma spectrometry was employed to quantify radionuclides from the 238U and 232Th decay series and 40K. The methodology followed IAEA protocols to ensure accuracy, including equilibrium sealing and extended counting durations. Results revealed significant differences in activity levels between samples: Ra-226 activity ranged from 279.01 Bq kg⁻¹ (Wizów) to 408.46 Bq kg⁻¹ (Wiślinka), while Th-232 was highest in Wizów (45.36 Bq kg⁻¹), suggesting variable radiological behavior. ICP-MS and XRF determined elemental composition, detecting REEs in all samples, with Police PG showing notably higher radioactivity, and Wizów PG containing approximately 0.09 wt% total REEs. This research demonstrates the viability of PG as a REE-bearing material and potential input for nuclear-related applications. The study’s primary objective is to assess PG as a dual-purpose material, addressing environmental waste management and the sustainable supply of critical resources for the nuclear fuel cycle. The work reduces the environmental impact and carbon footprint of REE extraction by identifying PG as an alternative to traditional mining. This approach aligns with circular economy principles in the nuclear sector and advances long-term sustainability goals, including reducing CO₂ emissions.

This research presents the application of Data-Driven Reduced Order Modelling (DDROM) to the DYNASTY experimental facility, addressing computational limitations of high-fidelity nuclear reactor models. The work demonstrates practical validation of theoretical ROM methods using real experimental data from a natural circulation loop.

Glioblastoma multiforme (GBM) is one of the most
aggressive and treatment-resistant brain tumors,
characterized by rapid proliferation, diffuse infiltration, and poor
prognosis. Despite current standards of care—surgical resection
followed by radiotherapy or chemotherapy. A major obstacle to
treatment success is the presence of hypoxic tumor niches that
harbor cells with metastatic phenotypes and intrinsic resistance to
conventional radiation. These regions are poorly oxygenated,
genetically unstable, and less responsive to
therapies, highlighting the urgent need for new strategies that
overcome radio resistance and engage anti-tumor immunity.
In this project, we investigated proton therapy as a precision
modality capable of delivering high linear energy transfer (LET)
radiation with minimal damage to surrounding healthy tissue.
Unlike conventional X-rays, proton beams deposit most of their
energy at the Bragg peak, enabling spatially localized dose
delivery. Beyond this physical advantage, we hypothesized that
proton therapy could stimulate immune responses through
activation of the cGAS–STING pathway, which detects cytosolic
double-stranded DNA and initiates the production of interferon-
beta (IFNβ), a key cytokine in anti-tumor immunity.
Using dose–response data from the U251 GBM cell line, we
incorporated IFN-β dose response into the FLUKA Monte Carlo
simulation framework. Simulations showed that a dose of
approximately 17 Gy optimized IFN-β induction without
triggering TREX1, a DNA exonuclease known to suppress immune
signaling at higher doses. The biological dose was modulated to
selectively cover a virtual subregion within the GTV, suggesting
the potential to stimulate localized immune responses in
treatment-resistant tumor microenvironment. These findings
support the dual role of proton therapy in delivering precise
radiation while also enhancing immunogenicity, offering a
promising foundation for future biologically guided radiotherapy
strategies in GBM. Further validation through in vitro
experiments and integration with immunotherapy could help
translate this approach into clinical protocols and expand its
relevance beyond GBM to other poorly immunogenic tumors

Purpose: Hypoxia in colorectal cancer (CRC) and gastroesophageal cancer (GEC) decreases
tumour responsiveness to radio and chemotherapy leading to cancer progression and poor
prognosis. This is the first study to utilise [18F]FAZA hypoxia radiotracer in patients with
CRC and GEC.
Methods: Six patients (mean age 68±8 years, 2 males and 4 females) with CRC and 4 patients
diagnosed with GEC (mean age 65 years, 3 males and 1 female) were included in the study.
[18F]FAZA was synthesised at the John Mallard Scottish PET Centre. After injection with
370 MBq of [18F]FAZA, PET/CT images with 60 min dynamic scan were acquired. In
addition, 15 min static scans 2 hr post injection were performed. 3D PET images were
reconstructed iteratively using an ordered subset expectation maximization (OSEM) method
and fused to the corresponding low-dose CT images. [18F]FAZA uptake parameters including
maximum standard uptake value (SUVmax), tumour-to-muscle ratio (T/M), tumour-to-bowel
ratio (T/B) and volume of interest (VOI) were measured.
Results: 4 out 6 patients with CRC (66%) showed clear uptake of [18F]FAZA in the primary
tumour. The mean tumour SUVmax was 2.2±0.91 (range 1.12 - 3.71). The tumour SUVmax
was significantly higher compared with muscle and bowel (t(5) =3.11, P=0.03), (t(5) =3.08,
P=0.03), respectively. However, tumour SUVmean didn’t differ significantly compared with
muscle and bowel (t(5) =2.41 , P=0.06), (t(5) =2.46 , P=0.06) respectively. The mean tumour
to muscle ratio (T/M) ratio was 1.89±0.64 (range 1.10 - 2.87), while the mean tumour to
normal bowel (T/B) was 1.92±0.64 (range 1.08 - 2.74). However, [18F]FAZA did not
accumulate in any of the tumours found in patients with GEC.
Conclusions: [18F]FAZA PET/CT imaging is suitable and feasible for detecting CRC hypoxic
tumour regions with image quality that can be used in clinical practice.

The challenges of food safety affect over 600 million people globally. In the rural parts of Africa, about 70% of the staple foods are toxic, containing cyanogenic glycosides, which can cause long-term health issues if improperly processed. Some of these foods include cocoyam and sorghum. To enhance sustainable food safety, routine quality control of these food plants is needed.
Previously, scientists used the wet chemical method for quality control, but it is destructive and time-consuming. This study employed neutron reflection- a rapid, environmentally sustainable, and non-destructive technique to determine the hydrogen content of various food samples. The experimental setup comprised a source holder, an Americium-Beryllium neutron source, a Helium-3 neutron detector, and a neutron counter. To establish a reliable calibration, liquid hydrocarbons served as reference standards, and the reflection parameter in these standards was measured using the neutron attenuation principle. This was used to determine the hydrogen content in four different varieties of cocoyam and sorghum. The data obtained showed a range of 6.1 ± 1.1 to 8.29 ± 0.10 hydrogen wt% of the respective food samples. Furthermore, we employed the alkaline picrate method to measure the cyanide content in the food samples. This method revealed a cyanide content range of 3.02 ± 0.07 to 6.8 ± 1.0 mg·L⁻¹.
Overall, the results showed an inverse relationship between the hydrogen content and cyanide concentrations, indicating a trend that can be used to assess cyanogenic food plants. This implies that neutron reflection can be used as a quality control tool for routine hydrogen assessment in cyanogenic food plants, underscoring the application of nuclear particle radiation in promoting food safety and sustainable development of the food industry.

The Red Palm Weevil (Rhynchophorus ferrugineus) is a destructive pest threatening palm cultivation globally, causing severe economic and ecological impacts. Conventional control methods, including chemical pesticides, have proven inadequate, prompting exploration of alternative, environmentally sustainable strategies. This study investigates the potential of X-ray radiation as a control method targeting R. ferrugineus larvae, focusing on its efficacy, physiological effects, and integration into pest management programs.

Controlled laboratory experiments were conducted to expose larvae at various developmental instars to calibrated doses of X-ray and gamma radiation. The effects were evaluated through larval mortality, behavioral observations, developmental progress, and reproductive outcomes. The results demonstrated a strong dose-dependent response. High radiation doses achieved complete larval mortality, while sublethal doses significantly altered behavior, reduced feeding activity, and impaired movement. Notably, surviving larvae that reached adulthood exhibited reduced fecundity and fertility, indicating potential for sterilization and long-term population suppression.

Radiation sensitivity varied by developmental stage, with early instars displaying greater susceptibility to lower doses. Sublethal exposures also induced developmental delays and morphological abnormalities, limiting reproductive success. These findings highlight the potential use of X-ray radiation in Sterile Insect Technique (SIT) applications and its broader utility in integrated pest management (IPM) systems.

In contrast to chemical pesticides, radiation-based control offers targeted action with minimal environmental impact, aligning with sustainability goals. However, further studies are necessary to assess ecological implications, optimize dosage, and explore practical implementation, including scalable field application and portable irradiation systems.
Overall, the study supports the viability of X-ray radiation as an effective, eco-friendly method for controlling R. ferrugineus larvae and underscores the importance of continued research to advance its operational use in palm protection strategies.

Optimizing the reload pattern of a nuclear reactor core, such as that of the APR-1400, is a complex and computationally intensive task due to the high dimensionality and non-linearity of the search space. These problems are characterized by numerous local minima, making it challenging to locate the global optimum using conventional deterministic methods. Metaheuristic algorithms like Simulated Annealing (SA) offer a promising alternative by probabilistically exploring the solution space. SA mimics the physical annealing process in metallurgy: it begins with a high "temperature" that encourages broad exploration, including the acceptance of suboptimal solutions, and gradually reduces the temperature to refine the search around promising areas.

In reactor core optimization, achieving a balance between competing objectives, economic performance (e.g., maximizing cycle length) and safety constraints (e.g., minimizing power peaking factor), further increases the problem complexity. Although SA provides flexibility and adaptability, its effectiveness can be hindered by the presence of deep local minima, which can trap the search process and prevent convergence to the global optimum. In this preliminary work, an SA framework is tailored for the Apr-1400 core reload problem to assess its feasibility.

Isotope 232Th is the only isotope of thorium naturally occurring in the nature and it is a fertile material. As a result of neutron capture by 232Th and two consequent beta decays appears fissile isotope 233U being a good nuclear fuel. Furthermore, the use of thorium may result in an extension of the fuel campaign.
There are numerous research papers available on the use of thorium in nuclear reactors. Unfortunately, so far, there has been no success in disseminating thorium-based fuel.
The authors investigated many geometrical models of fuel assemblies with ThO2, including Radkowsky thorium fuel seed-blanket concept or checkerboard layout. In this paper the impact on burnup of typical PWR fuel assemblies with addition of small amount of Thorium is reported.
The presented results are a small part of the research conducted on the effective use of thorium in PWR assemblies.

Following the Fukushima accident, the nuclear industry has accelerated the development of Accident Tolerant Fuel (ATF) technologies aimed at enhancing fuel safety and performance. This includes exploring advanced cladding materials such as chromium-coated Zircaloy, silicon carbide (SiC), and iron-based alloys like FeCrAl and stainless steel (SS), which offer improved resistance to high-temperature steam reactions that generate hydrogen.

In addition to cladding, fuel material innovation is also critical. Traditional uranium dioxide (UO₂) suffers from low thermal conductivity and limited heavy metal density, limiting its performance. While additives like beryllium oxide (BeO) can improve thermal conductivity, they reduce heavy metal content. Alternatively, fuels such as uranium silicide (U₃Si₂), uranium nitride (UN), and uranium carbide (UC) offer higher heavy metal density and better thermal properties, enabling extended cycle lengths without increased enrichment.

There is a strong interest in the United Arab Emirates (UAE) to extend APR-1400 reactor cycles from 18 to 24 months. This study evaluates U₃Si₂ as an ATF candidate for APR-1400, highlighting its 16.5% higher heavy metal content and favorable thermal conductivity behavior compared to UO₂. However, the increased initial reactivity from U₃Si₂ requires effective reactivity control measures. To this end, the High Gadolinium (HIGA) burnable absorber concept is investigated as a potential solution to suppress excess reactivity, reduce soluble boron requirements, and avoid issues like a positive moderator temperature coefficient.

Abstract
Controlling reactivity is still a major problem that is typically solved with burnable poisons (BPs). Gadolinium is one example of a single BP that can cause power peaking and uneven depletion. The use of hybrid burnable poisons (HBPs), which balance residual reactivity with neutron absorption, presents a possible substitute. Hybrid BP combinations in VVER-1200 cores have received little attention in previous research, which mostly focuses on conventional poisons (such as Gd alone). OpenMC and ORIGEN have been partially integrated for isotopic and decay analysis in VVER-1200 reactors and inadequate long-term forecasting of cycle life and k-eff degradation in hybrid BP scenarios. Fuel cycle economics and reactor safety are the driving forces behind the deployment of HBPs. Extended operational cycles, better power distribution, and less reactivity swings are all possible with HBPs. For effective fuel cycle planning, waste control, and economical operation, an accurate core lifespan prediction is essential. The performance of conventional burnable absorbers, such as gadolinium or boron compounds, may be improved by hybrid burnable poison (HBP) techniques that combine materials (e.g., GdO₃ + ErO₃). A thorough computational examination of a generic VVER-1200 core using hybrid burnable poisons is presented in this paper. Predicting isotopic evolution, core lifespan, and reactivity behavior (k-effective) under various burnable poison loading procedures is the goal of the study. To allow a thorough assessment of fuel cycle performance, a coupled OpenMC–ORIGEN process is created and used to simulate burnup and post-irradiation inventories.

A robust and accessible database of radioecological parameters is essential for effective emergency preparedness and response to radiological and nuclear incidents. Such a database compiles critical information on radionuclide behavior in various environmental compartments, including soil-to-plant transfer factors, distribution coefficients, ecological half-lives, and bioaccumulation rates across ecosystems.
We present here a database of such factors, summarizing literature accumulated over many years of work.

By providing a centralized repository of validated parameters, this database serves as key tool for radiological protection, public safety, decision support systems and environmental management.

In total, the database contains in total over
3300 entries for transfer factors for over
65 different nuclides from
36 different elements,
collecting data from 23 different countries.

This paper presents an innovative methodology for analyzing and modeling fission product releases in Small Modular Reactors (SMRs) using the Probabilistic Model Chain Methodology (PMCM). The proposed approach integrates detailed site-specific meteorological data, reactor design parameters, and severe accident scenarios to enable more accurate predictions of radioactive plume dispersion and dose assessment. The methodology supports both deterministic and probabilistic Level 3 Probabilistic Safety Assessments (PSAs), allowing for improved definition of emergency planning zones. This contributes to enhanced public safety, more targeted emergency preparedness, and greater cost efficiency in the deployment of SMRs.

The rapid detection and characterization of accidental radioactive releases is critical for public safety, environmental protection, and informed emergency response. In the Gulf region, the increasing presence of nuclear power infrastructure—including the operational Barakah Nuclear Power Plant (UAE), the Bushehr facility (Iran), and a planned plant in Saudi Arabia—necessitates robust regional monitoring and response systems. Accidental atmospheric releases from these facilities could pose transboundary risks, highlighting the importance of coordinated preparedness and decision-support capabilities.

This study presents a comprehensive framework tailored to the Gulf region that integrates three core components: (1) generation of a state-of-the-art dispersion scenario database using a source-receptor matrix approach; (2) spatial optimization of ground-based monitoring networks for enhanced early detection and plume tracking; and (3) application of inverse modeling techniques to reconstruct the source term, including emission rate and possible location. Together, these components aim to strengthen nuclear safety and emergency response capabilities under scenarios of severe atmospheric releases.

The dispersion of hazardous pollutants in urban environments presents a complex challenge, especially in the context of nuclear accidents or malicious releases involving radioactive substances. In such scenarios, understanding the transport mechanisms of contaminants is crucial for emergency preparedness, risk assessment, and public safety. Numerical simulations based on computational fluid dynamics (CFD) have become indispensable tools for predicting pollutant behavior in densely built areas. However, the accuracy and reliability of these predictions depend heavily on the choice of turbulence modeling approach.

This study evaluates the performance of three widely used modeling strategies - Reynolds-Averaged Navier-Stokes (RANS), Large Eddy Simulation (LES), and hybrid RANS–LES methods - for simulating the dispersion of hazardous plumes in complex urban geometries. The simulations were carried out using a realistic urban layout and pollutant source configuration based on the 'Michelstadt' wind tunnel experiment, conducted by Hamburg University as part of the COST Action ES1006. The selected scenario reflects situations relevant to nuclear safety, such as accidental releases near nuclear facilities or during the transport through densely populated regions.

The study revealed that the scale-resolving methods, such as LES and Hybrid, provides significantly improved accuracy in predicting pollutant dispersion in realistic urban settings. While URANS captures the general flow and dispersion trends and performs well with enhanced inflow conditions, it tends to underpredict lateral spreading and turbulence-driven mixing, especially in complex downstream regions. In contrast, the scale-resolving methods better resolve unsteady vortex shedding and turbulence structures, leading to more accurate estimates of streamwise and spanwise velocity profiles, and ultimately a more realistic prediction of pollutant accumulation and transport - particularly in recirculation zones and urban canyons. These findings emphasize the importance of turbulence model selection in CFD applications for urban air quality and nuclear emergency planning.

The application of advanced fluids with improved thermal properties is a potential approach to enhance heat transfer efficiency in nuclear reactors, resulting in an elevation in the safety margin. This study investigates the impact of a hybrid TiO₂-Cu/water nanofluid on the thermodynamic behavior of coolant flow within a triangular array of a PWR subchannel. A single-phase flow of a homogeneous mixture of species was assumed. The analysis considers Reynolds numbers ranging from 20,000 to 80,000, in increments of 20,000, and evaluates three nanoparticle volume ratios: 1% TiO₂–3% Cu, 2% TiO₂–2% Cu, and 3% TiO₂–1% Cu. Key performance parameters analyzed include variations in the Nusselt number, convective heat transfer coefficient, temperature distribution, enthalpy gain, shear stress, and pressure drop. This study exhibited that the addition of nanoparticles enhances the overall convective heat transfer coefficient compared to the base fluid (water). A higher volume fraction of TiO₂ improves heat transfer performance, although the rate of improvement relative to the water decreases as the Reynolds number increases. Among all the combinations, the 3% TiO₂–1% Cu ratio demonstrates the most efficient heat transfer based on temperature profile and Nusselt number trends. However, it also incurs a notable pressure drop. In contrast, the 1% TiO₂–3% Cu combination yields a reduction in pressure drop compared to the water. The flow dynamics and heat dissipation characteristics of the hybrid nanofluid were thoroughly examined using CFD simulations in ANSYS Fluent.

The development of Generation IV reactors to replace coal-fired power plants has recently gained global attention. The Sustainable Nuclear Energy (SNE) research team is currently designing a 300 MWth Molten Salt Reactor (MSR) for repowering coal power plants. As a Generation IV reactor, the MSR uses liquid fuel consisting of fluoride or chloride salt mixtures, allowing high-temperature operation at low pressure [1]. For safety, matching power and fuel flow distributions is crucial to avoid temperature hotspots. [2]. Assuming a uniform power distribution, the fluid flow must also be uniform.
Several previous studies [2], [3], and [4] have calculated the flow rate ratio of each channel relative to the average flow rate. Therefore, this study discusses how an alternative MSR–Repowering design with improved matching between fluid flow and power distribution in the core can be proposed.

Using Ansys Fluent 2024 R2 and OpenFOAM, this study successfully reduced flow deviation. The initial case showed Sreaktor and Sring values of 59.43% and 5.0%, respectively. To lower these values, modifications were introduced sequentially. The upper plenum adjustment reduced the deviation to 53.18% and 4.4% for Sreaktor and Sring, respectively. This was followed by an alternative lower plenum geometry, reducing the values further to 50.39% and 4.89%. The final configuration, with the addition of a shroud, achieved the most significant improvement, yielding Sreaktor and Sring values of 31.63% and 4.9%, respectively. Verification using OpenFOAM showed insignificant errors compared to Ansys Fluent, with deviations of less than 5%.

Phase change materials (PCMs) are widely applied in thermal energy storage systems due to their high latent heat capacity, and understanding their solidification behavior is crucial for optimizing heat release performance. In this study, the solidification process of a selected PCM was numerically investigated using STAR-CCM+ to analyze the associated flow characteristics and phase transition dynamics. To validate the simulation results, experimental measurements of the solidification process in a vertical annular container were compared with numerical predictions. The close agreement between the temperature evolution at different measurement points confirmed the accuracy of the model. Based on the validated simulation, detailed analyses of temperature distribution, solid fraction evolution, and flow field development during solidification were conducted. The results indicate that solidification initiates near the cooled surface, progressing inward with a non-uniform front due to the combined effects of conduction and natural convection in the remaining liquid phase. As the solidification front advances, natural convection weakens, leading to a transition toward conduction-dominated heat transfer in the later stages. The findings provide deeper insight into the interplay between flow behavior and heat transfer during PCM solidification, which can guide the design of more efficient thermal energy storage systems.

The main aims of this work was to determine the effects of ionize radiation (gamma and e-beam rays) on starch: PVA and Starch: PVA:CNC films. 3 different starches where used for this purpose; pre-irradiated corn and wheat starches and native wheat starch. The results obtained from this works shows that in terms of mechanical test, corn Starch had highest elongation at break whereas pre-irradiated wheat Starch films had the highest tensile strength. Moreover, the addition of CNC greatly improved the mechanical test and the used of ionizing radiation increases the mechanical test and reduces the swelling behaviour and gel fraction due to degradation caused by radiation. The TGA analysis show that they was crosslinking forms when these films were irradiated as different doses and as conclusion we can say that irradiation greatly improved the quality of these films

The TNG-40 is a cutting-edge 40 MWe mobile pressurized water reactor (PWR) designed as a floating small modular reactor (SMR) to support Saudi Arabia’s energy diversification goals. Its compact and efficient design, featuring UO₂-Silumin composite fuel and optimized enrichment strategies, enables safe, reliable, and flexible power generation for maritime logistics, emergency coastal power, and dual civil-military applications. The multilayer radiation shielding, inspired by proven designs like the KLT-40S and NS SAVANNAH reactors, is rigorously modeled using OpenMC Monte Carlo simulations to ensure safety while minimizing material use.
A novel closed-loop integration strategy combines high-fidelity Monte Carlo simulations, Artificial Neural Network (ANN) surrogate modeling, and Non-dominated Sorting Genetic Algorithm II (NSGA-II) multi-objective optimization. This approach balances computational efficiency and accuracy by iteratively refining shielding designs based on surrogate predictions validated through additional simulations. The TNG-40’s modularity and multi-purpose capabilities, including electricity generation, industrial steam, and desalination, position it as a versatile solution for future energy and industrial needs, adhering to international safety standards and operational reliability.

With sea freight accounting for around 90% of global trade, screening shipping-container cargo presents significant challenges at seaports, particularly in countries like Bangladesh. Now a days Nuclear and Radiological application in Bangladesh is quite expanding. For this reason, it has become an alarming issue to ensure nuclear and radiological security in Bangladesh to prevent illicit trafficking of radioactive material. Main seaport of Bangladesh is Chittagong seaport which uses conventional x-ray to scan and detect any illegal material to stop illicit trafficking of radioactive material. In modern countries neutron imaging is introduced as a more effective material scanning process in the basis of identifying concealed radioactive sources. To enhance detection capabilities, Neutron Imaging can be integrated with X-ray imaging in a multi-modal system. While X-rays efficiently identify anomalies in cargo, NR excels in penetrating dense materials like metals and plastics that may obstruct X-ray scans. The aim of this study is to analyze the scope of neutron imaging in the context of Bangladesh. The focus of this study is on Chittagong Sea Port, where the possibility of integration of neutron imaging technology is discussed based on its prospects. The international standards and codes along with economic impact are also considered. This study will pave the way for consideration about installing neutron imaging technology alongside conventional x-ray imaging. The result of this study will depict and represent the effectiveness of neutron imaging techniques in Bangladesh which may motivate other scientists, engineers and regulatory bodies to step forward about installing Neutron imaging techniques in Bangladesh with more extensive research and study.

The modeled detector has been used to examine the versatility of the MCNP package in modeling the response of high pressure ion chambers to photon radiation. The results show that the response characteristics of a gas ion chamber filled with xenon are similar to the measured results from other studies of similar ion chambers. The model results indicate that while the use of a compressed xenon gas ion chamber detector improves the low energy response of the detector. Other detection gases like krypton, argon, neon, or helium can be easily modeled using the MCNP model created for this work, as was done for one part of this investigation.

Compact spherical fusion tokamaks require robust neutron shielding to protect superconducting cores from radiation-induced degradation within tight spatial constraints (~50 cm). Tungsten carbide (WC) is a prime candidate shielding material due to its excellent neutron and gamma attenuation properties, but its response to neutron irradiation is not fully understood. This study examined WC's irradiation-induced lattice swelling, defect evolution, and thermal transport degradation using tungsten ion irradiation (0.13–13 dpa; 100–400 °C). Grazing Incidence X-ray Diffraction showed significant initial lattice expansion (1.3% at 0.13 dpa, 100 °C), decreasing with higher temperatures and doses, ultimately transitioning to lattice contraction at high dose (13 dpa). Transmission Electron Microscopy and FIB-SEM revealed pronounced grain boundary cracking in coarse-grained WC and enhanced resistance in fine-grained WC due to refined microstructure. Preliminary Transient Grating Spectroscopy results demonstrated a dramatic reduction in thermal diffusivity (order-of-magnitude drop at lowest dose), highlighting substantial microstructural damage. This work informs WC shielding optimisation and future irradiation studies for advanced tokamak designs.

The interaction of ionizing radiation with living organisms and their components is a fundamental issue in the field of radiation protection. DNA, as the carrier of genetic information that is sensitive to mutations, and for this reason, understanding the nature of the impact of ionizing radiation on DNA is particularly important. The scope of presented work was applying two spectroscopic methods, ultraviolet spectroscopy and circular dichroism spectroscopy to analyze the changes in DNA, with focus on its conformation, that undergo in effect of low doses irradiation with β and γ rays.

This paper aims to investigate the compact crossflow heat exchangers (HXs) and their implementation in the lead fast reactors (LFRs). Two different analyses have been performed, following the same geometrical parameters in terms of tube diameter and pitch-to-diameter ratios, as reported in the research work conducted by Ciurluini et al. [1], which addresses the same reactor concept proposed by Newcleo.
The first regards a 2D numerical domain, focused on characterising the Nusselt number in such compact configurations, comparing the CFD standalone results with the different values obtained by the correlations available in literature.
The second case, which is explicated in this extended abstract, is a lead flow field extracted by the cylindrical layout SG, constituted by 72 banks of Archimedean spiral tubes, and is analysed employing STH/CFD code coupling approaches, to take advantage of both codes’ best features.

The study was conducted to explore the effects of lanthanum oxide (La₂O₃) on the physical, mechanical, and nuclear shielding properties of six Boro-Silicate glass compositions (S1 to S6), with La₂O₃ concentrations ranging from 10 to 35 mol%. Mechanical properties, including Young’s modulus, bulk, shear, longitudinal moduli, microhardness, and Poisson’s ratio, were evaluated using the Makishima–Mackenzie model. Results show that increasing La₂O₃ content enhances the glass density and elastic properties, while slightly reducing microhardness due to increased network rigidity. Gamma-ray shielding characteristics were investigated across the 0.015–15 MeV energy range using both the MCNP5 simulation code and XCOM software. Key parameters assessed included mass attenuation coefficient (µm), half-value layer (HVL), effective atomic number (Zeff), radiation protection efficiency (RPE), and buildup factors (EBF and EABF). The S6 composition (45B₂O₃–10SiO₂–10CaO–35La₂O₃) exhibited the best performance, achieving the highest µm, Zeff, and RPE, and the lowest HVL, EBF, and EABF offering 100% X-ray attenuation between 10–150 keV and excellent gamma shielding at higher energies. Neutron shielding capability was evaluated through the macroscopic effective removal cross-section (ΣR), with S1 showing the highest value and S6 performing comparably to conventional materials like graphite, concrete, and basalt magnetite. Additionally, the mass stopping power (MSP) and projected range (PR) of protons and alpha particles were analyzed using the SRIM code across 0.01–15 MeV. The S6 glass again demonstrated superior performance, with the lowest MSP and shortest particle ranges. Overall, these results confirm that La₂O₃-doped Boro-Silicate glasses, particularly the S6 composition, are highly effective for shielding against X/gamma rays, neutrons, and heavy ions. Their versatility makes them suitable for use in medical diagnostics, nuclear waste containment, and radiation protection infrastructure.

This study focuses on enhancing safety validation methods for small modular reactors (SMRs) by investigating the behavior of passive containment cooling systems (PCCS) during severe accidents. Many SMR designs rely on a containment-water pool coupling, which acts as a final heat sink during events like loss-of-coolant accidents (LOCA). The thermal-hydraulic behavior inside the containment—such as gas stratification, temperature distribution, and natural convection—is influenced by complex interactions between steam, hydrogen, and the cooling pool.

To better understand these dynamics, researchers are using both numerical simulations and technical-scale experiments. In this work, the GOTHIC code is applied to simulate SMR-relevant accident scenarios, which are then validated through experiments at the THAI+ test facility. The facility includes interconnected pressure vessels, and it is being expanded to incorporate a new SMR-type vessel capable of simulating various containment geometries and conditions. This vessel can operate at pressures up to 64 bar and can be fully or partially submerged in water.

Extensive instrumentation, including thermocouples and flow meters, enables high-fidelity monitoring of key parameters like temperature, pressure, and condensation. In test scenarios, steam is injected into the vessel, initially condensing on the inner walls before reaching steady conditions. Hydrogen is later added to study gas stratification and condensation dynamics.

Preliminary simulations show the formation of distinct stratified layers—steam, hydrogen, and condensate—within the containment, offering crucial data for validating thermal-hydraulic models. The facility supports a "double-blind benchmark" environment, providing unprecedented experimental conditions for testing and validating LP and CFD simulations of next-generation reactor safety systems.

The hybrid reactor can be considered an attractive actinide burner for transmuting transuranic nuclear waste and reproducing fissile material. Its availability of high-energy neutrons from an external source coupled to the system (such as neutrons produced by spallation reactions or fusion reactions) makes it possible to burn transuranics even in a subcritical operating mode, making such systems inherently safe. Several proposals for hybrid reactors have been studied. However, the most developed concepts are the hybrid fusion-fission reactors (FFH) and the Accelerator Driven Systems (ADS). One of the major obstacles to the construction of these reactors is the development of materials capable of withstanding the hostile environments to which they will be exposed due to the high temperatures and intense bombardment of high-energy particles, conditions created by the technology of the external neutron source. In addition, these machines are designed as large projects, consequently increasing their cost and construction time. Considering overcoming such obstacles, some SMR (Small Modular Reactor) projects present designs favorable for adaptation to a small subcritical hybrid reactor project, such as the Swedish Advanced Lead Reactor (SEALER), a small modular fast neutron nuclear reactor. The lead-cooled core composition and geometry allow for the maintenance of a hardened neutron spectrum, which is necessary for the transmutation by fission of transuranics in reprocessed fuel. Thus, this work aims to evaluate the burnup of reprocessed fuels in a small subcritical hybrid reactor based on SEALER, utilizing a fusion neutron spectrum characteristic of an FFH with plasma parameters obtained from the tokamak reactor ITER. Aspects such as the distribution of the reprocessed fuel in the core, which maximizes the transmutation rate, and the analysis of the performance of the external neutron source will be investigated.

If my submission is accepted, I would like to respectfully request consideration for one of the 20 available grants to support travel and accommodation.

The Effect of Control Rod Position On Control Rod Worth In Gama Micro Space Reactor

Nuclear energy is one of the most promising energy sources to meet the world's growing energy demand. This calls for the development of more efficient and safer Generation IV reactor technologies. The FUJI-U3 Molten Salt Reactor (MSR), as one of the Generation IV reactor designs, offers several advantages, including high fuel efficiency, long-term reduction of radioactive waste, and the ability to operate at low pressure. However, the neutronic design of this reactor involves high complexity due to multiple operational parameters. This study aims to develop a predictive model using Artificial Neural Network (ANN) to accelerate the calculation process of the effective multiplication factor (k_(eff⁡)) based on Monte Carlo simulation data generated by the OpenMC computer program. By varying the Th/U ratio, operating temperature, and core geometry parameters, a dataset consisting of 400 configurations was obtained. Prior to training the ANN model, the simulation data were analyzed and normalized. The ANN model was developed using a three-hidden-layer architecture with the GELU activation function and optimized using the Adam algorithm. Model evaluation on test data yielded a Mean Absolute Error (MAE) of 0.00228, Mean Squared Error (MSE) of 0.00001, Root Mean Squared Error (RMSE) of 0.00273, and a coefficient of determination (R²) of 0.99626, indicating a very high level of prediction accuracy. These results demonstrate that the ANN model is effective and can be used as a predictive tool in the design and optimization of the FUJI-U3 MSR.

The safe and sustainable management of radioactive waste across all phases — including conditioning, storage, transportation, and disposal — requires reliable, autonomous, and long-term monitoring solutions. This work presents a novel integrated system developed by the University of Pisa within the H2020 PREDIS project, aiming to address the operational challenges posed by sensor heterogeneity, regulatory compliance, and long-term data management. The architecture combines a Django-based metadata framework for structured representation of facilities, waste packages, and measurement records, with ThingsBoard for IoT device provisioning, telemetry handling, and real-time data processing. Wireless sensor nodes, equipped with solid-state micro-power gamma and thermal neutron detectors, perform continuous in-situ measurements directly from cemented waste drums. Data are autonomously transmitted via long-range LoRaWAN connectivity, ensuring robust wireless coverage over extended distances, even within complex storage and transportation environments.

A key element of the system is its sustainability-oriented design, which incorporates ultra-low power electronics combined with energy harvesting solutions to enable long-term autonomous operation, minimizing maintenance and infrastructure requirements. The middleware layer provides secure communication, metadata enrichment, and ensures full traceability of monitoring data throughout the entire lifecycle of radioactive waste. The proposed modular and scalable architecture lays the foundation for integration into future frameworks such as EURAD-2 (Work Package 5 - ICARUS), supporting not only long-term monitoring but also enabling advanced functionalities such as predictive analytics, anomaly detection, and automated decision-making, ultimately reinforcing the safety, sustainability, and resilience of radioactive waste management infrastructures.

The integration of digital Instrumentation and Control (I&C) systems in nuclear facilities creates complex cyber-physical-human systems, introducing vulnerabilities at the confluence of nuclear safety, physical security, and cybersecurity. Conventional risk assessment methodologies often analyze these domains independently and are thus inadequate for evaluating blended threats that exploit such systemic interdependencies. This paper proposes a two-phase analytical framework to address this deficiency. The first phase utilizes Systems-Theoretic Process Analysis (STPA), a top-down hazard analysis technique, to identify Unsafe Control Actions (UCAs) that can precipitate a loss event. This identification is agnostic to the causal factor, encompassing component failure, human error, or malicious cyber-physical attacks. In the second phase, high-consequence scenarios identified via STPA are modeled using Dynamic Probabilistic Risk Assessment (DPRA). DPRA is a simulation-based method that quantifies risk by modeling the temporal evolution of event sequences and dynamic system interactions. The framework's application is demonstrated through a case study of a blended threat against a model research reactor. The analysis reveals systemic vulnerabilities, for example where manipulated sensor data could cause operators to perform incorrect actions. The results quantify how such threats degrade system resilience by reducing the time available for appropriate operator response, thereby increasing the conditional probability of adverse outcomes. This integrated, systems-theoretic approach provides a more robust and quantitatively rigorous framework for the assessment and management of risk in complex, modernized nuclear facilities.

This study investigates the thermal behaviour of deep geological repositories (DGRs) designed to isolate high-level radioactive waste (HLW) and spent nuclear fuel (SNF) in Bangladesh. Given the long-term heat generation from radioactive decay, understanding temperature evolution within the repository is crucial for safety and design optimization. Using ANSYS software, this study simulates the heat transfer process from waste canisters into surrounding engineered and geological barriers, focusing especially on bentonite buffers and different regional clay types found in Bangladesh. The research compares two types of heat source models: line heat source and volumetric heat generation. It also evaluates various initial decay power levels to reflect realistic reactor waste scenarios. The study incorporates regional clay data such as smectite, Illite, and kaolinite; to assess their thermal performance under repository conditions. The key findings show that both the type of clay and the waste’s initial heat output strongly influence the surrounding repository temperature by about 50–100°C and 10–20°C, respectively. Illite-kaolinite-rich regions show better thermal insulation, whereas Smectite results in higher peak temperature of 324°C. Overall, this study offers a region-specific thermal assessment for nuclear waste disposal in Bangladesh. It highlights how geological variation influences repository safety and provides guidance for future repository site selection and engineering design. The work will contribute to the long-term nuclear waste management strategy of Bangladesh and emphasize the importance of localized modelling in nuclear safety assessments.

This paper presents an extension of the Incompressible Schrödinger Flow (ISF) method for computational analysis of inviscid fluid flows. The ISF method represents an alternative computational approach to traditional Computational Fluid Dynamics (CFD) and spectral methods for inviscid fluid analysis. The fundamental concept exploits the mathematical analogy between hydrodynamics and quantum mechanics, which was initially proposed by Madelung in 1927. This approach transforms the non-linear hyperbolic Euler equations into a linear complex partial differential equation based on the Schrödinger equation. This research contributes to developing quantum-inspired computational methods for fluid dynamics, offering potential advantages in terms of computational efficiency and numerical stability for specific classes of inviscid flow problems.

The current work aims to provide a complete and explicit physics-based model, within the uncertainty of the experimental data, for Vertical upward annular two-phase flow. The model will only require data that is known before-hand without any restriction on its range.

The evolving landscape of nuclear energy marked by the deployment of small modular reactors (SMRs), the development of advanced fuel cycles and rising geopolitical tensions is fundamentally reshaping the requirements for nuclear safeguards. No longer limited to ensuring regulatory compliance, safeguards now demand a highly competent, adaptive, and technologically literate workforce. This shift calls for a new paradigm in education and training: one that fosters critical thinking, interdisciplinary collaboration, and a deep understanding of emerging verification technologies.
Special attention is given to the role of the European Nuclear Education Network (ENEN) as one of the main catalysers of this transformation. By integrating academic excellence with hands-on training and international collaboration, ENEN serves as a bridge between educational institutions and real-world nuclear safeguards practice. Such instrumentation such as the ENEN2plus mobility programme, nuclear safeguards theoretical and hands-on practical courses, scientific competitions and academic programmes are key enablers of this competence-based shift driven by ENEN.
This paper explores how safeguards education is transitioning from a compliance-focused model to a competence-oriented approach and examines the implications of this shift for workforce development in the nuclear sector. Diverse audiences from inspectors and engineers to regulators, researchers, legal and policy advisors require customized learning pathways that reflect their roles in safeguarding nuclear materials in increasingly complex environments.
The paper concludes with practical recommendations on how educational strategies can be aligned with the demands of the nuclear renaissance, particularly in areas such as digital safeguards (e.g., tomographic techniques, AI-assisted material accountancy or surveillance, etc.), small modular reactors (SMR) technologies evolving and strategic capacity-building in politically sensitive regions. This evolution in safeguards education is not only enhancing professional readiness but is also helping to ensure the long-term resilience and credibility of the global non-proliferation regime.

The paper presents IAEA NPP simulators for education and training as well as related considerations for gap analysis and simulator choice.

More than twenty years of IAEA experience in the distribution and application of basic education and training simulators shows that education and training simulators are an effective medium for knowledge transfer.

The IAEA regularly arranges for the development of simulation software and corresponding training materials, sponsors training courses and workshops, and distributes computer programs and documentation.

The University of Leeds is proudly home to one of the largest nuclear engineering groups within the UK, where multiphase flows are instrumental in much of the work we do. Research at Leeds encompasses the whole fuel cycle, from characterizing bubbly reactor flows and investigation of new molten salt designs to intensified separators and the transportation of nuclear wastes. Additionally, with the UK’s push to renew its nuclear power enabling its ambitious Net Zero goals, this is a critical time for nuclear engineering. To support our nuclear mission, both for new nuclear reactors builds and to ensure the safe disposal of legacy wastes, Leeds has established the Multiphase Fluid Flow In Nuclear systems (MULTIForm) suite, as part of the UK’s National Nuclear Users Facility (NNUF). Here, we will introduce the MULTIForm facility and how it supports the next generation of nuclear researchers.

The global pursuit of net-zero carbon emissions by 2050 necessitates a comprehensive transition to low-carbon energy sources. To achieve this target, a rapid and extensive transition in energy supply is necessary, leveraging all available technologies. Alongside renewable energies such as solar and wind, nuclear energy is considered an important element in reducing carbon emissions.
We present a detailed Life Cycle Assessment (LCA) of the Barakah Nuclear Energy Plant in the United Arab Emirates (UAE), benefiting from the unique opportunity to utilize detailed input data collected during the construction and operation phase of the newly deployed plant in the Middle East and North Africa (MENA).
The LCA study demonstrates the low carbon emission of nuclear energy and confirms the success and important role of the UAE's nuclear program in achieving the net-zero goal.

This study presents the development and validation of a detailed neutronic model of the Open Pool Australian Light water (OPAL) research reactor using OpenMC, an open-source Monte Carlo particle transport code. The OPAL reactor is a 20 MW pool-type facility utilizing low-enriched uranium (LEU) U₃Si₂-Al dispersion fuel assemblies and equipped with five control rods. The reactor serves multiple purposes, including radioisotope production, materials testing, and neutron beam research.

This validated model provides a reliable computational framework for future high-fidelity studies involving control rod worth, power distributions, and burnup analysis. It also demonstrates the capability of OpenMC for full-core modeling of complex pool-type research reactors, supporting its application in core design, safety analysis, and performance evaluation.

Abstract
The Fukushima Daiichi accident in 2011 exposed how zirconium alloy clad fuel can oxidize rapidly at high temperatures and react violently with coolant, producing hydrogen and ultimately leading to hydrogen explosions. This event spurred the development of Accident Tolerant Fuel (ATF) for Light Water Reactors (LWRs). Iron based alloys (e.g., FeCrAl, APMT™, 310SS, 304SS) offer enhanced thermo mechanical strength and much lower steam oxidation reactivity, thereby reducing hydrogen generation; however, they suffer a reactivity penalty because their neutron absorption cross sections are roughly fifteen times greater than those of zirconium alloys. Previous studies have proposed multi-parameter mitigation strategies, including reduced cladding thickness, reduced pellet diameter, and increased uranium enrichment, to offset this penalty.
In the present work, a preliminary thermal-hydraulic investigation was carried out to quantify the effects of alternative cladding materials and reduced cladding thickness. A three-dimensional model of a single subchannel within a fuel assembly (based on the BEAVRS benchmark geometry) was developed, and ANSYS Fluent was employed to simulate steady-state coolant flow and heat transfer under nominal operating conditions. The thermal performance of FeCrAl, APMT™, 310SS, and 304SS claddings was evaluated against that of conventional Zircaloy-4.
Results show that all materials exhibit similar temperature distribution trends, differing only in magnitude according to their thermal properties. Reducing cladding thickness from 0.5715 mm to 0.350 mm produced an approximate 7 K rise in outer-surface temperature across all alloys and increased the Nusselt number by about 10.6.
Further studies are needed to assess the impact of thinner iron-based claddings on Departure from Nucleate Boiling (DNB) and Critical Heat Flux (CHF) margins during design-basis transients, as well as to evaluate high-temperature creep and embrittlement behavior. Subsequent Loss-of-Coolant Accident (LOCA) and Reactivity-Initiated Accident (RIA) simulations will be required to ensure compliance with IAEA standards and best practices.

The Near Surface Disposal (NSD) facility which designed by the Center for Radioactive Waste Technology (CRWT) needs to be evaluated from the possibility of radionuclide release from the facility to ensure its radiation safety to human and environment. Cesium-137 was chosen as the radionuclide of concern due to its high mobility in environment, long half-life, and total activity in waste inventory at CRWT.  Radiological impact (i.e dose and excess cancer risk) was assessed using RESRAD-OFFSITE 4.0 version. Input data such as dimension and characteristics of NSD and local soil, activity of inventory, weather data, food and water consumption, occupancy time of critical grup, and other inputs related to the impact to critical groups had been prepared. Sensitivity analysis on thickness of compacted bentonite from 0.05 m to 1 m was conducted. The purpose of this study is to estimate the potential radiological impact of an NSD facility by considering the effect of bentonite thickness variation on NSD foundation. From the results, maximum dose of critical group still below 1 mSv/year. Increasing the bentonite thickness from 0.05 m to 1 m will reduce the critical group dose from 3.38×10-2, 2.99×10-2, 2.04×10-2, and 1.35×10-2 mSv/y, while its cancer risk from 2,12×10-4 to 1,5×10-4. Changes in the thickness of the compacted bentonite layer are able to reduce the dose and risk effectively. It means that the thicker compacted bentonite layer can reduce the impact of Cs-137 radiological hazards from waste packages stored in the NSD facility to critical groups and the environment.

As shown in Fukushima Daiichi accidents, radioactivity absorption in water ponds (pool scrubbing) is highly efficient in mitigating Source Term to the environment during Severe Accidents (SA). Therefore, SA analytical tools should include an accurate modelling that provides reliable insights into pool scrubbing effects and reduces the overall uncertainties in fission products released to the environment. The last goal of this PhD project will aim to make a visible improvement of safety analysis of nuclear power plants. All the potential approximations to modelling Source Term attenuation in pools will be explored, from correlations to mechanistic modelling both in 0D and 3D approach and enhanced to the best of their capabilities to develop a definitive sound modelling. Such a development will be thoroughly validated and its uncertainties properly characterized in risk-relevant accident scenarios.

The application of advanced fluids with improved thermal properties is a potential approach to enhance heat transfer efficiency in nuclear reactors, resulting in an elevation in the safety margin. This study investigates the impact of a hybrid TiO₂-Cu/water nanofluid on the thermodynamic behavior of coolant flow within a triangular array of a PWR subchannel. A single-phase flow of a homogeneous mixture of species was assumed. The analysis considers Reynolds numbers ranging from 20,000 to 80,000, in increments of 20,000, and evaluates three nanoparticle volume ratios: 1% TiO₂–3% Cu, 2% TiO₂–2% Cu, and 3% TiO₂–1% Cu. Key performance parameters analyzed include variations in the Nusselt number, convective heat transfer coefficient, temperature distribution, enthalpy gain, shear stress, and pressure drop. This study exhibited that the addition of nanoparticles enhances the overall convective heat transfer coefficient by an average of 5.2% compared to the base fluid (water). A higher volume fraction of TiO₂ improves heat transfer performance, although the rate of improvement relative to the water decreases as the Reynolds number increases. Among all the combinations, the 3% TiO₂–1% Cu ratio demonstrates the most efficient heat transfer based on temperature profile and Nusselt number trends. However, it also incurs a notable pressure drop. In contrast, the 1% TiO₂–3% Cu combination yields a reduction of 4.31% on average in pressure drop compared to the water. The flow dynamics and heat dissipation characteristics of the hybrid nanofluid were thoroughly examined using CFD simulations in ANSYS Fluent.

This the first draft of the poster for team-G ,(Numerical Prediction of Liquid Metal Thermal-Hydraulics in Backward Facing Step for Advanced Nuclear Reactor Design)

this our summer training project with JSI and we will continue working on it and improve it

I. Introduction
II. Nuclear-Energy Steam Grid
II.A. Steam Grid for Heavy Industry
III. Major Result
IV. Conclusions
Acknowledgments
References

• This research aims to model a landmine detection system using Prompt
  Gamma Neutron Activation Analysis (PGNAA), which identifies
  explosives by detecting gamma emissions from key elements such as
  hydrogen, carbon, nitrogen, and oxygen.
• Gamma signature peaks from these four elements were used to validate
  the model.
• The presence of hydrogen, carbon, and nitrogen peaks can be used to
  confirm the existence of explosive landmines, enhancing the
  reliability and flexibility of our system compared to previous
  systems that rely solely on the nitrogen peak.
• The simulation focused on optimizing moderator and shielding
  parameters and evaluating the system's detection capabilities.
  Performance analysis was conducted by varying factors such as sample
  composition, casing type, burial depth, and soil type.
• The results confirmed the system’s ability to detect explosives even
  when encased in plastic or metallic casings and buried in soil.

In light of the escalating climate crisis, the global dependence on fossil, non-renewable, and polluting energy sources presents a significant challenge to mitigating the adverse environmental and climatic effects resulting from the unchecked use of such sources. In this context, nuclear energy emerges as a clean and environmentally friendly alternative, with the potential to meet the growing energy demand efficiently, on par with more conventional energy sources. Small Modular Reactors (SMRs) stand out as promising contenders in the global nuclear sector, offering advantages such as operational flexibility, robust safety measures, and the ability to meet local energy demands. These reactors hold substantial potential for shaping the future of nuclear energy generation. The SMART (System-integrated Modular Advanced Reactor) is a small modular reactor, moderated and cooled by pressurized water, developed by the Korea Atomic Energy Research Institute (KAERI), with a thermal power output capacity of up to 330 MWt. It has already been licensed, and two units are set to be constructed in Saudi Arabia in the near future. The objectives of the present work are to model the main components of the SMART reactor using the RELAP5 MOD 3.3 thermal-hydraulic analysis code, widely employed in the licensing stages of power and research reactors, and verifying the modeling under steady-state operating conditions and evaluating the reactor's behavior during transient situations. The results from steady-state simulations were compared to reference data and showed good agreement with the expected values, within the acceptable error margins found in literature. The analysis of pressure drop in the primary system, as well as the temperatures of fuel, gap, cladding, and coolant, yielded consistent results within the expected ranges. Additionally, the results obtained from transient simulations are also presented and discussed in this work, providing a comprehensive view of the reactor's behavior under dynamic operating conditions.

This study investigates the feasibility of Boron Neutron Capture Synovectomy (BNCS) as a targeted treatment for rheumatoid arthritis using accelerator-based neutron sources. BNCS offers a non-invasive alternative to surgery by delivering high-linear energy transfer (LET) radiation directly to diseased synovial tissue through the 10B(n,α)7Li reaction. A compact epithermal neutron beam was modeled using the 7Li(p,n)7Be reaction at 1.97 MeV, with an optimized moderator and reflector configuration. A detailed knee phantom was constructed, including tissue-equivalent layers such as skin, synovium, cartilage, and bone. Monte Carlo simulations were performed using MCNPX to estimate neutron interaction and dose deposition in the synovium under various boron concentrations. The results demonstrated efficient dose delivery to the target region with minimal exposure to surrounding healthy tissues. These findings support the potential clinical applicability of BNCS, highlighting its effectiveness, safety, and adaptability for treating joint conditions in an outpatient setting.

The Kingdom of Morocco has a 2 MW research reactor (TRIGA MARK II) and an interim storage facility at the Maamoura Nuclear Research Center (CENM), managed by the Nation-al Center for Nuclear Energy, Science, and Technology (CNESTEN), designed for radioac-tive waste management and treatment on a national scale. Spent radioactive resins (SRR) are commonly used to clean and treat water pool reactors in the nuclear industry. Cement binders are considered the major matrices for the solidification/encapsulation of these radioactive wastes. Previous work from our laboratory has shown that blended cement incorporating clays can potentially yield high-performance cement binders for encapsulation of SRR at a pilot scale (A.Sadiq et al., 2021). Taking into account these endeavors, this study aimed to investigate the effect of introducing clay into the cemented radioactive resin matrix and how it affected the amount of spent resin supported by the cementing formulation. Various amounts of SRR loading ranging from 20 wt% (wet base) within the cement binders were performed. The results show that the resin ratio increased by 140.9% compared to the formu-lation adopted by CNESTEN and by 11.11% compared to the formulation adopted by previ-ous studies (A.Sadiq et al., 2022). The compressive strength of the matrix was improved after 28 days of curing time. These results suggest that the developed cementitious binders incor-porating clay are a potential matrix for the solidification of these radioactive wastes.

poster for the reasearch paper

Abstract
Flow pulsations in tightly spaced fuel rod bundles are an important phenomenon in reactor thermal-hydraulics, as they can cause vibrations that may compromise structural integrity. These pulsations, originating from coherent gap vortex streets between fuel rods, can induce vibrations and fatigue in reactor components. In this work, we investigate this behavior using Computational Fluid Dynamics (CFD) simulations based on Reynolds-Averaged Navier–Stokes (RANS) models within the ANSYS Student Version. The Shear Stress Transport (SST) k–ω turbulence model is applied to capture unsteady flow characteristics in a rod bundle geometry inspired by the Hooper experiment, which features a pitch-to diameter ratio of 1.107. The geometry and flow parameters are simplified and scaled to remain within the limits of available computational resources, while preserving the essential physics of axial flow pulsations. While high-fidelity methods such as DNS or LES provide more accurate representations of turbulence, they are computationally demanding, so the focus is on testing the capability of RANS turbulence models to reproduce the essential features of axial flow pulsations. Velocity data from the simulations are analyzed in both the time and frequency domains to identify characteristic pulsation patterns. In particular, time histories of the gap-center velocity are examined using power spectral density (PSD) to extract the dominant pulsation frequency, which is then compared with values reported in the literature. The project highlights both the strengths and the limitations of RANS-based CFD in capturing this phenomenon and provides a useful baseline for future work with more advanced methods

At King Fahd University of Petroleum and Minerals (KFUPM), the radiological safety of a proposed shallow-land repository—projected to be reinforced-concrete vaults measuring 24 m × 18 m × 9 m for low- and intermediate-level radioactive waste (LILW)—was evaluated with the RESRAD-OFFSITE v4.0 code alongside specific hydrogeological, geochemical, and lifestyle information. The inventory was obtained from the Gyeongju LILW facility in South Korea and was adjusted to the estimated waste volume at KFUPM. The dose constraint adopted for compliance was 0.1 mSv/year, as recommended by IAEA for the individual exposed maximally. Baseline calculations predict the single dose maximum of approximately 11,000 years will be 9 × 10⁻⁴ mSv/year, dominated by the activation product Ni-59. This peak only represents 0.9% of the regulatory limit. Bounding sensitivity tests, produced by multiplying every release fraction uniformly by 0.1 and 10, yielded peak doses of 1 x 10⁻⁶ mSv/year and 3 x 10⁻³ mSv/year respectively, both remaining safely below the compliance limit of 0.1 mSv/year. The region near the closure has short-lived Co-60 and Co-58 dominating contribution, with Ni-59, Nb-94, and Pu-239 taking over after 90,000 years. This concludes that the repository does not exceed the exposure criteria set under normal conditions, which plus other best case scenario uncertainties would make him more comfortable against captivity for solubility, sorption, and release fraction, remains stable against plausible uncertainties in disposition. The study justifies the acute radio-toxicity surveillance in the facility disaster program while also showing the South Korea policy can be used in foreign regions.

Monte Carlo methods have become indispensable tools in the analysis and safety assessment of nuclear power plants and nuclear facilities. Their ability to model complex particle transport processes with high spatial and energy resolution enables advanced applications across reactor neutronics, inventory and activation analysis, radiation protection, reactor safety and environmental monitoring.

This paper outlines recent applications of Monte Carlo simulations in five critical areas. The authors show, that with a basic model, once commenced in the design phase of the reactor, the whole lifespan of the reactor can be supported in its various stages from design to build through operation into decommissioning.

The effects of helium ion (He+2) implantation into glassy carbon (GC) were systematically investigated. He+2 ions with an energy range of 17 keV were implanted into GC to fluences of 10¹⁶, 10¹⁷, and 10¹⁸ cm-² at room temperature (RT). The as-implanted GC samples were subsequently vacuum annealed at 300 °c, 500  °c, and 800  °c for 1 hour. The structural evolution of GC was characterized using Raman spectroscopy and transmission electron microscopy (TEM). A fluence-dependent trend in displacement per atom (dpa) and He concentration was observed. Raman spectroscopy revealed progressive structural disorder and amorphization at fluences of 10¹⁷ and 10¹⁸ cm-², marked by merging and redshifts of the D and G peaks, indicating tensile strain in the carbon matrix. Partial recovery of D/G peak separation and crystalline order was observed, especially at 800 °c for the 10¹⁶ cm-² fluence. TEM micrographs showed a confined damaged region of about 130 nm, with distinct defect aggregation towards the bulk for fluences of 10¹⁶ and 10¹⁷ cm-², whereas the defect aggregation appeared in two channels for the fluence of 10¹⁸ cm-². At a fluence of 10¹⁷ cm-², nonlinear dispersion and saturation effects were observed. Overall, annealing facilitated partial microstructural recovery, particularly for samples with fluences of 10¹⁶ and 10¹⁷ cm-² at 800 °c.

I. Introduction
The goal of this work is to improve the understanding of gas build-up mechanisms and support the design of safer containment and transport systems in nuclear environments. In particular, the focus is on hydrogen accumulation, which poses a significant risk due to its flammability and potential for explosion. Developing and validating accurate predictive models is essential for ensuring safety in such scenarios.
II. Experimental Approach
To support the development and validation of these models, a series of small-scale explosion experiments will be conducted using the MK-II test rig. This experimental setup features a high-pressure spherical chamber equipped with four impellers, which enables control over both laminar and turbulent flow conditions. Hydrogen will be introduced into the chamber through a controlled source and ignited using an integrated ignition system. The vessel is designed to withstand internal pressures of up to 40 bar.High-speed cameras will be used to visually capture the ignition and flame propagation processes. At the same time, pressure sensors located at various positions inside the chamber will record the pressure dynamics throughout each test. These measurements will provide detailed insight into the behaviour of hydrogen under different flow and ignition conditions.

Abstract

This research focuses on the design and analysis of a High-Temperature Gas-Cooled Reactor (HTGR), specifically utilizing the High Temperature Test Reactor (HTTR) model, with the objective of achieving an outlet temperature of 850 °C and a thermal power output of at least 70 MWth without altering the core volume. The study meticulously addresses critical safety aspects, emphasizing the reactor's inherent safety features and compliance with rigorous evaluation criteria. Neutronic behavior is investigated using Open-MC simulations to assess the effective multiplication factor (K_eff) and the impact of various control rod configurations on reactor stability. Additionally, thermal hydraulics are examined, focusing on helium’s properties as a coolant, including mass density and heat transfer characteristics, with detailed analyses of temperature distributions within the reactor core to optimize thermal management and prevent overheating. This work enhances our understanding of high-temperature reactor designs and their operational efficiencies, laying the groundwork for future advancements in nuclear reactor technology.

New types of nuclear reactors are pushing the limits of what fuels need to deliver in terms of safety, efficiency, and long-term performance. This report looks closely at two important fuels that are leading the way: High-Assay Low-Enriched Uranium (HALEU) and Tristructural-Isotropic (TRISO) fuel. HALEU, with uranium-235 enrichment levels between 5% and 20%, makes it possible to build smaller, longer-lasting reactor cores and improves neutron economy. These features are especially useful for small modular reactors, microreactors, and fast reactors. TRISO fuel, built with multilayer coatings around uranium kernels, is known for keeping fission products trapped even under very high temperatures, sometimes reaching over 1600°C. This makes it a strong candidate for high-temperature gas-cooled reactors and systems that focus on accident tolerance. While a lot of progress has been made, the studies also point to some challenges, like gaps in the HALEU supply chain and the difficulty of making TRISO fuel at scale. Fixing these problems with more investment, smarter regulations, and new innovations will be key to making sure these fuels help nuclear energy stay strong and sustainable in the future.

I. Introduction
Nuclear waste generates hydrogen at a slow rate via various mechanisms and this hydrogen must be released in a safe manner. In some instances, the hydrogen is released in discrete “burps” rather than as a continuous release. Ignition from these releases has the potential to injure people nearby. The goal of this work is to improve the understanding of overpressure development when small releases of hydrogen ignite [1]. Developing and validating accurate predictive models is essential for ensuring safety in such scenarios.
II. Research Summary
To support the development and validation of these models, a series of small-scale explosion experiments will be conducted using the MK-II test rig. This experimental setup features a high-pressure spherical chamber equipped with four impellers, which enables control over both laminar and turbulent flow conditions. Hydrogen will be introduced into the chamber through a controlled source and ignited using an integrated ignition system. The vessel is designed to withstand internal pressures of up to 40 bar. A photo and Schematic of the MK-II test rig is shown in Fig. 1 below [2,4].

High-speed cameras will be used to visually capture the ignition and flame propagation processes. At the same time, pressure sensors located at various positions inside the chamber will record the pressure dynamics throughout each test. These measurements will provide detailed insight into the behaviour of hydrogen under different flow and ignition conditions [2-4].
The experimental data collected from the MK-II rig will play a critical role in validating computational models of hydrogen explosions. These models aim to accurately simulate flame propagation and pressure development within confined geometries. The findings will not only contribute to more reliable simulations but also offer deeper insight into hydrogen combustion phenomena relevant to nuclear safety.

The tin-tailing processing industry in Malaysia has long operated with minimal regulatory oversight, particularly regarding radiological safety and environmental management. This has resulted in significant exposure risks to workers and surrounding ecosystems due to elevated level of Naturally Occurring Radioactive Materials (NORM) and heavy metals. Radiological Impact Assessment (RIA) conducted on environmental sample revealed concentrations of radionuclides such as 226Ra, 232Th and 40K ranging between 0.1-10.0, 0.0-25.7, and 0.1-5.8 Bq/g, respectively. These levels contributed to annual effective doses (AEDs) exceeding the 1 mSv/y limit recommended by UNSCEAR and enforced by AELB. Correspondingly, the calculated radium equivalent activity (Raeq) indicates potential gamma-ray hazards to human health. Heavy metal pollution indices also demonstrated substantial contamination, particularly arsenic (As) and iron (Fe), with average exposure through soil ingestion and dermal contact exceeding tolerable thresholds, raising both non-carcinogenic and carcinogenic health concerns for industry workers. Statistical correlations between NORM and trace elements further highlight the compounded environmental burden. The current situation is exacerbated by the 1994 exemption order, which excluded this industry from compliance with the Atomic Energy Licensing Act 1984. Despite clear evidence of radiological of radiological harm, regulatory response and industry accountability remain limited. Lessons from other countries show that strict enforcement, comprehensive regulations, and punitive measures are effective in mitigating such risks. This study underscores the urgent need for Malaysia to reform its nuclear safety framework to safeguard public health and environmental sustainability in the face of ongoing tin-tailing activities.

Submission for the Saudi International Conference On Nuclear Power Engineering

This research presents a parametric extension of Dynamic Mode Decomposition (pDMD) applied to Molten Salt Fast Reactor dynamics analysis. The work addresses the computational limitations of Full-Order Models in nuclear reactor system evaluation, where discretized partial differential equations provide high accuracy at significant computational cost.This work contributes to the broader development of reduced-order modeling techniques for nuclear reactor analysis, addressing the need for computationally efficient models suitable for multi-query applications, design optimization, and real-time monitoring systems.

This two-page abstract presents a benchmarking study of the OpenMC Monte Carlo code using the University of Florida Training Reactor (UFTR) as a graphite-moderated test case. It outlines the development of a high-fidelity OpenMC model, the simulation setup, and comparison with experimental reactivity data. The abstract demonstrates OpenMC’s accuracy in modeling graphite-moderated systems and discusses its potential for Generation IV reactor analysis.

The use of Covalent Organic Frameworks (COFs) in uranium adsorption has gained attention due to their potential in nuclear waste remediation. Traditional solvothermal synthesis methods, while effective, require high temperatures, long reaction times, and toxic solvents, limiting their scalability. This study introduces an alternative Dissolution-Precipitation (DP) synthesis method for COF formation, exploring the impact of different alcohol precipitants (ethanol, isopropanol, and butanol) on COF morphology, stability, and uranium adsorption performance. The novelty of this work lies in evaluating the role of alcohol choice in tailoring COF properties, addressing a research gap in solvent engineering for COF development.

COFs were synthesized using both DP and solvothermal methods, with yields of 87% (COF-EtOH), 93% (COF-IPA), and 47% (COF-BuOH) for the DP method, and 98% for solvothermal COFs. Structural and morphological analyses were performed using NMR, FTIR, SEM, BET, and XRD. Uranium adsorption was assessed under varying conditions, with COF-IPA showing the highest adsorption capacity of 800 mg/g, achieving 80% uranium uptake. Langmuir modeling (R² > 0.99) confirmed monolayer chemisorption, with COF-IPA outperforming COF-EtOH (625 mg/g) and COF-BuOH (714 mg/g) due to its optimal porosity, sulfonate density, and colloidal stability.

The findings highlight that alcohol selection significantly influences COF performance, with COF-IPA showing the most promise for uranium recovery. This study demonstrates that DP synthesis, with careful solvent choice, provides an environmentally friendly and scalable route for COF-based uranium capture, positioning it as a viable alternative to traditional methods in nuclear waste remediation.

Efficient treatment of radioactive wastewater containing toxic heavy metals such as Pb²⁺ and Cu²⁺ remains a pressing environmental challenge. This study explores the application of an Agitated Tubular Reactor (ATR) as an intensified process alternative to conventional batch systems. Using a pilot-scale Coflore® ATR, the performance of clinoptilolite and co-precipitation agents under varying agitation conditions (0–5 Hz) for simultaneous removal of Pb²⁺ and Cu²⁺ at 200 ppm concentrations was evaluated. Comparative experiments demonstrated that the ATR achieved removal efficiencies of 98% for Pb²⁺ and 97% for Cu²⁺, outperforming natural clinoptilolite alone (58% Pb²⁺, 67% Cu²⁺) and matching or exceeding batch systems. Enhanced lateral mixing through the ATR’s sinusoidal agitation promoted superior mass transfer, dispersion, and floc formation, resulting in compact sediment structures with compressive yield stress comparable to those from batch processes. Dye tracing confirmed efficient pseudo-plug flow behavior and shear-driven mixing, particularly at a frequency of 5 Hz. Unlike batch reactors, which are limited by operational intermittency and clogging, the ATR offers continuous operation, reduced fouling, and higher space-time yields. Moreover, its modular, compact design enables scalability and mobile deployment, making it highly suitable for on-site applications at nuclear facilities. Given its superior performance, process intensification capabilities, and operational flexibility, the ATR represents a viable and promising future solution for treating radioactive effluents containing heavy metals in both industrial and environmental remediation contexts.

The treatment of radioactive waste effluents has become increasingly critical with the global expansion of nuclear energy. Conventional methods, such as ion exchange, evaporation, and chemical precipitation, are effective but often limited by high energy requirements and secondary waste generation. Polyamide-based nanofiltration (NF) membranes offer a promising alternative due to their high selectivity, lower operational pressure, and scalability. However, these membranes exhibit vulnerability to gamma irradiation present in low- and intermediate-level radioactive waste (LILW), leading to structural degradation, reduced salt rejection, and diminished water permeability.
This review synthesizes recent advancements in the development of radiation-resistant polyamide NF membranes for nuclear wastewater treatment. Key strategies include surface functionalization with polyethyleneimine (PEI), the incorporation of Prussian Blue and graphene oxide, and the spray coating of titania nanosheets. These modifications have achieved enhanced rejection rates for radionuclides such as Co²⁺, Sr²⁺, and Cs⁺, with values reaching up to 99.5% while preserving or improving water flux. Furthermore, uranium removal efficiencies have demonstrated strong pH and ionic dependency, with rejection rates ranging from 4% to 98%.
Despite these improvements, challenges remain in addressing fouling, maintaining long-term structural integrity under radiation exposure, and achieving consistent rejection of monovalent radionuclides. Future directions emphasize the need for hybrid membranes that integrate adsorption and separation functions, as well as real wastewater validations under operational conditions.
In conclusion, polyamide NF membranes hold significant potential as core components in next-generation radioactive effluent treatment systems, aligning with the goals of sustainable and safe nuclear energy deployment.

This research applies Marchetti's logistic substitution theory to analyze competition between fossil fuels and low-carbon energy sources using global energy consumption data spanning 1965-2023. The study consolidates nine energy sources into two competing categories: fossil fuels (coal, oil, natural gas) and renewables plus nuclear power (solar, wind, hydropower, nuclear, biofuels, other renewables). The research demonstrates that logistic substitution models retain descriptive power for macroscopic energy system analysis.

The Saudi Arabia ambitious decarbonization plan targets achieving net-zero greenhouse gas emissions by 2060. Among different kinds of carbon-free energy sources, nuclear energy has the potential to play an essential role in the decarbonization plan, due to its considerable reliability and stability, and its high energy density making it suitable for large scale energy needs. The two energy sectors, water desalination and air conditioning, represent a significant share of overall energy consumption in Gulf Council Cooperation Council countries including Saudi Arabia. Currently, these two sectors are covered by electric-driven technologies, putting a significant pressure on power grid, and leading to a huge seasonal electricity consumption gap, between summer and winter. This study proposes and evaluates the idea of employing Heat-only Small Modular Reactors, such as TEPLATOR, integrated with thermally driven water desalination and district cooling systems and energy storage. A typical system is proposed for a typical case study namely the Line project in Saudi Arabia, and a systematic approach is developed for its design and energy management optimization. This idea is to be techno-economically evaluated, as one of the solutions for decarbonizing water desalination and air conditioning energy needs.

Jordan's energy strategy is centered on establishing a sustainable energy system by diversifying its energy sources and maximizing the use of indigenous resources. This strategy reflects Jordan's commitment to achieving carbon neutrality and contributing to global efforts to limit global warming to below +1.5°C. To meet these objectives, the country is prioritizing the promotion of renewable energy, enhancing energy efficiency, and reducing reliance on imported fossil fuels. The ultimate aim is to ensure a reliable, sustainable, and environmentally friendly energy supply, aligning with global climate change mitigation efforts.
While Jordan acknowledges the crucial role of renewable energy in achieving a sustainable future, it also recognizes the current limitations of renewables in fully supporting the national energy grid. To address this, the country is exploring additional methods to decarbonize both the electric and non-electric sectors as part of its broader strategy for sustainable, low-carbon economic growth.
In response to these challenges, Jordan launched a 10-year National Energy Sector Strategy in 2020, which serves as a key component of its long-term, low-emission development plan. The strategy aims to reduce carbon emissions by 10% by 2030 and improve the energy mix. It emphasizes not only the expansion of renewable energy sources but also the enhancement of energy efficiency and the adoption of other low-carbon technologies. Through this approach, Jordan aims to achieve its climate goals while fostering economic growth and strengthening energy security. In a statement, Jordan emphasized its intention to leverage all available technologies to boost local energy dependency.

Saudi Arabia's steel industry is undergoing significant expansion, with plans to increase crude steel production from approximately 10 million tons today to 20 million tons by 2030 in pursuit of self-sufficiency. This growth aligns with the industrialization goals of Vision 2030 and the national commitment to achieving net-zero emissions by 2060, necessitating the decarbonization of steelmaking. Currently, the sector relies heavily on fossil fuels, primarily natural gas for Direct Reduced Iron (DRI) and oil/gas-fired power for Electric Arc Furnaces (EAF), resulting in estimated annual CO₂ emissions of 10 to14 million tons.

This study explores the potential of High-Temperature Gas-cooled Reactors (HTGRs) to enable green steel production by 2060, through the cogeneration of carbon-free electricity and high-grade heat for hydrogen production. A hydrogen-based DRI-EAF route is proposed as the pathway to achieving net-zero steel, with the integration of advanced HTGRs (≥900 °C outlet) into Saudi steel plants. A mid-term hybrid scenario (2025-2035) deploying current-generation HTGRs (~300 MWe / 750 MWt) serves as a transitional approach, while the long-term focus shifts toward next-generation HTGRs for a fully hydrogen-driven process by 2060.

Our findings indicate that HTGR-SMRs are optimally suited for this application. A single 300 MWe / 750 MWt HTGR can deliver approximately 5 TWh of electricity and 10 TWh of high-temperature heat annually, sufficient to support the production of around 2 million tons of steel via conventional DRI–EAF. Alternatively, if its output is dedicated to high-temperature electrolysis, the reactor can produce up to 210,000 tons of green hydrogen per year, enabling approximately 3.8 million tons of green steel through a 100% hydrogen-based DRI process. In both cases, substantial CO₂ emissions reductions can be achieved, positioning HTGR-SMRs as a strategic enabler of Saudi Arabia’s industrial decarbonization goals.

Saudi Arabia’s transition from internal combustion vehicles (ICVs) to electric vehicles (EVs) is a key strategy for reducing greenhouse gas (GHG) emissions and achieving the nation’s Vision 2030 sustainability goals. However, the environmental benefits of EV adoption depend critically on the carbon intensity of the electricity used for charging. This study employs a comparative Life Cycle Assessment (LCA) using OpenLCA and the ReCiPe 2016 Midpoint (H) method to evaluate the global warming potential of four scenarios: ICVs, EVs powered by fossil fuels, EVs powered by a 50/50 mix of fossil fuels and nuclear power, and EVs powered exclusively by nuclear power. The analysis is based on a functional unit of 150,000 km traveled by subcompact vehicles, accounting for engine and battery deterioration over a five-year period. Results indicate that ICVs produce the highest CO₂ emissions (69,500 kg CO₂ eq), followed by EVs powered by fossil fuels (24,400 kg CO₂ eq), EVs with mixed electricity sources (15,600 kg CO₂ eq), and EVs powered solely by nuclear energy (6,920 kg CO₂ eq). Transitioning from ICVs to EVs reduces CO₂ emissions by 64.9%, while integrating 50% nuclear power into the grid achieves a 77.5% reduction, and a fully nuclear-powered EV scenario yields a 90.05% reduction. Nuclear power plants (NPPs) thus play a pivotal role, offering a 71.65% emissions reduction compared to fossil-fuel-powered EVs. The findings underscore that coupling EV adoption with a decarbonized electricity supply—particularly via nuclear energy—is essential for maximizing climate benefits in Saudi Arabia’s transport sector. This study provides quantitative evidence to support policy decisions on sustainable energy and transportation transitions in the Kingdom.

The current study is interested in investigating the radiological behavior and mineralogical controls of natural radioisotopes (²³⁸U, ²²⁶Ra, ²³²Th, and ⁴⁰K) in granitic pegmatites from the Abu Zawal Area (AZA), Eastern Desert, Egypt. The pegmatites, enriched with thorite, zircon, monazite, ferrocolumbite, and fergusonite, display abnormally high activity concentrations up to 1272 Bq/kg for ²³⁸U, 723 Bq/kg for ²³²Th, and 570 Bq/kg for ²²⁶Ra far exceeding international safety limits. Radiological assessments revealed elevated radium equivalent activity (Raeq ≤ 1984 Bq/kg), hazard indices (Hex ≤ 5.36), and excess lifetime cancer risk (ELCR ≤ 20.9 × 10⁻³), all surpassing recommended thresholds by UNSCEAR and ICRP. Absorbed dose rates ranged from 182 to 978 μSv/year, with annual effective dose equivalents showing wide spatial variability linked to Th-rich mineral zones. Mineralogical analysis identified phase-specific controls on radiological behavior: thorite and monazite drive thorium-related gamma emission and radon/thoron release, while zircon and fergusonite are responsible for uranium enrichment and decay chain disequilibrium. Even low-activity minerals like ferrocolumbite can contribute to localized radiation hotspots through co-enrichment in U and Th. These results highlight critical relationships between specific mineral phases and their associated radiological risks, offering a threefold contribution: i) Establishing a baseline for regional radioactivity mapping and U/Th enrichment; ii) Supporting the safe industrial use of pegmatitic materials by identifying high-risk phases; iii) Informing mining operations and processing strategies through mineralogically targeted radiation protection, including ventilation and shielding protocols. The study underscores the need for integrated radiological and mineralogical assessments in evaluating natural radiation hazards in granitic pegmatites and contributes to safer practices in resource utilization and environmental health protection.

This study presents the development and evaluation of novel chromium–niobium (Cr–Nb) alloy coatings as an accident-tolerant fuel (ATF) solution for Boiling Water Reactors (BWRs). Cr–Nb coatings with 15 and 25 at.% Nb were deposited via magnetron sputtering on Zr-based cladding and subjected to out-of-pile testing. Long-term dynamic autoclave corrosion tests exceeding 120 days under BWR Normal Water Chemistry (NWC) showed stable performance with no coating degradation, cracking, or delamination. In contrast to pure Cr coatings, which gradually dissolved, Cr–Nb coatings remained chemically stable. High-temperature steam oxidation at 1100 °C demonstrated that Cr–Nb coatings effectively delayed oxidation for up to approx. 45-60 minutes before protectivity was lost. SEM/EDS analysis revealed the formation of distinct Cr- and Nb-rich oxide phases, indicating separate oxidation behavior and limited interdiffusion. This layered oxide structure likely contributes to early-stage protectivity. These results highlight the potential of Cr–Nb coatings as a viable ATF option for BWRs, supporting further evaluation under irradiation and accident conditions.

In this study, gamma radiation capability of glass systems made up of a mixture of boron and coconut shell ash (CSA) doped with varying concentrations of tellurium-dioxide (TeO2) (BTC1–BTC5) were investigated with the composition 58CSA–(42–x)B2O3–xTeO2 (where x = 0.1, 0.2, 0.3, 0.4, and 0.5 mole%). The glass samples were fabricated via the melt-quenching technique at 1000oC melting temperature. The mass attenuation coefficient (MAC), linear attenuation coefficient (LAC), half value layer (HVL), tenth value layer (TVL), mean free path (MFP), and effective atomic number (Zeff) were determine across different photon energies (ranging from 0.01 to 10 MeV) using WinXcom computer software. For the MAC, a sharp absorption edge at 0.05 MeV is observed due to the presence of photoelectric effect. As photon energy increases, all results of the fabricated glass samples exhibit a rapid decrease for the all radiation shielding parameters determined, indicating changes in the glass systems as they tend to lose their shielding properties. Sample BTC5 which have higher density (approximately 2.176 g/cm³) with higher content of TeO₂, was found to have the highest MAC and LAC, lowest HVL, TVL, and MFP, and the largest effective atomic number across all energy bands. As such, sample BTC5 consistently demonstrates superior shielding performance. For the observed plotted spectra of the fabricated samples, it affirms that increasing TeO₂ content improves the gamma shielding performance of the fabricated glass systems. This finding highlight the potential of coconut shell ash (CSA) as an eco friendly, low-cost heavy element alternative in glass composites for radiation shielding performance.

Five tin-bismuth alloys doped Sb element with composition Sn -50Bi and (92-x)Sn -8Sb-xBi (x= 10.0, 20.0, 30.0 and 40.0 wt.%) was prepared by melting pure Sn, Bi and Sb at 500 °C then casting at 30 °C. The phase composition, morphology, microhardness, and mechanical behavior of the prepared alloys were characterized using X-ray diffraction (XRD), scanning electron microscopy (SEM), Vicker microhardness and tensile test technique respectively. The structure analysis of alloys is composed of matrix phase of tetragonal β-Sn, rhombohedral α-Bi, rhombohedral Sb rhombohedral SbSn and new rhombohedral Bi0.92Sb0.08 intermetallic compounds (IMCs). The results show that the suppression of the brittle SbSn as well as appearance of new Bi0.92Sb0.08 IMC in addition to crystallite size refinement due to the addition of Sb. Therefore, the creep resistance, hardness and mechanical properties of the prepared alloys are improved. It has been observed that the addition of Bi up to 40 wt.% can significantly affect the properties of Sn-8Sb alloys. Furthermore, these alloys' ability to shield against gamma radiation was investigated and evaluated in the energy range of 0.015 to 15 MeV using MCNP5 code and WinXCom software. A number of factors are computed in order to fully understand the researched alloy's radiation and neutron shielding properties. The findings showed that raising the Bi concentration improves radiation shielding qualities. The Sn-50Bi alloy demonstrated the highest shielding performance compared to other prepared alloys, common shielding materials, and recently studied alloys. The Sn-50Bi alloy had the best neutron attenuation capability. The Sn-50Bi alloy demonstrated the best attenuation performance for protons and alpha particles, making it a potential material for radiation shielding applications in industry, medicine, and nuclear waste storage.

The Kingdom of Saudi Arabia has joined the distinguished group of nations benefiting from nuclear reactor technology. Given the unique environmental conditions of the region, the most effective strategy for nuclear reactor deployment is the development of low-power micro-nuclear plants that can be rapidly installed in remote locations [1]. The key considerations in nuclear reactor implementation include innovative design, operational safety, and economic feasibility[2,3]. Accordingly, this research focuses on the development of a low-power micro-nuclear reactor. In the initial phase, a technical and economic assessment of small modular reactors and micro-nuclear reactors is conducted to evaluate cost-effectiveness and safety performance. The second phase involves the development of passive safety features to enhance the safety performance of the selected micro-reactor. Subsequently, the conceptual design of the reactor is developed based on neutronic and thermal-hydraulic calculations. Computational tools are primarily utilized for thermal-hydraulic analysis to ensure accurate performance assessment. This study provides a comprehensive preliminary evaluation of micro-reactors, paving the way for further research and benchmark studies on advanced nuclear reactor technologies. Additionally, a key objective of this project is to train graduate students in nuclear engineering, equipping them with expertise in reactor benchmarking and the development of innovative reactor designs.

A brief consideration is given of nuclear waste disposal options as the end stage of nuclear waste management (NWM) activities. Nuclear (the same as radioactive) waste is waste that contains, or is contaminated with, radionuclides, at activity concentrations greater than clearance levels set by the national regulatory organisations with international advice provided by the International Atomic Energy Agency (IAEA). Nuclear waste results as a byproduct of various nuclear energy applications ranging from medicine to power generation, as well as from processing of materials containing naturally occurring radionuclides (i.e. NORM) such as those within oil and gas production, ore beneficiation, and water purification. NWM envisions all administrative and operational activities involved in the handling, pretreatment, treatment, conditioning, transport, storage and disposal of radioactive waste. Advantages provided by the borehole disposal both for vertical and horizontal options are due to the placement of nuclear waste packages straightly into boreholes (tunnels, wells) via the near surface reception sections of GDFs. Instead, mined GDFs utilise accessing disposal wells after transportation of nuclear waste packages into the disposal section of repository located deeply underground. This requires significant infrastructure deep underground which results in high maintenance costs and actual or potential upgrading expenditures .

Microbiologically influenced corrosion (MIC) in spent nuclear fuel storage pools presents a critical challenge to the structural integrity of nuclear containment systems, with significant implications for radioactive material safety and long-term waste management. This review combines bibliometric analysis and microbiological insights to explore global research trends from 2013 to 2024, identifying 42 relevant publications through a refined Scopus database search using keywords aligned with the INIS Thesaurus. Analytical tools such as VOSviewer and Biblioshiny revealed evolving research priorities centered on microbial activity, biofilm formation, and corrosion mechanisms, with keyword co-occurrence maps underscoring the prominence of terms like "microbiology," "biofilm," and "copper canister." Detection methods for corrosive microbes include molecular techniques (PCR, qPCR, next-generation sequencing), electrochemical approaches (electrochemical impedance spectroscopy), and advanced imaging (scanning electron microscopy, atomic force microscopy), enabling detailed characterization of dominant microbial groups such as Chloroflexi and Proteobacteria across various storage surfaces. Conventional mitigation strategies—such as cathodic protection and protective coatings—are being complemented by innovative approaches like quorum sensing inhibition, engineered microbes, and biodegradable inhibitors, though these novel methods require further field validation. By mapping research output, identifying key microbial contributors, and evaluating emerging detection and mitigation technologies, this study provides a valuable resource for researchers, engineers, and policymakers aiming to enhance monitoring systems, develop effective countermeasures, and ensure the long-term safety and environmental sustainability of spent fuel storage infrastructure.

Handling, transporting, and storing spent nuclear fuel (SNF) present enduring challenges from more than 70 years of nuclear power operations. Transportation and storage systems are rigorously evaluated for criticality, shielding, heat, and structural integrity under conservative conditions, optimized through system analysis. The load conditions for assessing the structural safety of SNF casks are categorized as normal, off-normal, accident, and natural phenomena. Extensive tests of SNF containers to simulate 99% of all travel-related accidents were conducted, while test data acquisition was mainly used to develop and validate sophisticated computer models. The tests also considered further severe impacts, confirming the package's ability to contain the SNF and the accuracy of the computer analyses. The most severe design-based accidents involving transportation casks typically include a 9-meter drop onto a rigid surface, considering various impact orientations. In the context of storage facilities, an aircraft impact is regarded as one of the most significant risks during the SNF storage phase, despite the low probability of occurrence. These two severe cases encompass a wide range of impact loads. Efforts have been made to develop methods for modeling and simulating these two cases within a safety assessment framework. This involves addressing varied impact conditions and considering the characteristics of SNF and their detailed responses [1-10].

This research provides an overview of the primary methods proposed, assumptions related to numerical and modeling, and impact conditions developed by the author. Finally, it highlights knowledge gaps and areas that require further investigation.

This study evaluates the synthesis and performance of polyphosphate-functionalized mesoporous silica nanoparticles (MSNs) for the adsorption of uranium (U) from aqueous media. The nanoparticles were synthesized via a modified sol-gel method and functionalized using O-phosphorylethanolamine. Characterization via SEM, BET, FTIR, and TGA confirmed successful modification.

Adsorption experiments were conducted to assess the influence of pH, temperature, contact time, adsorbent dose, and competing ions. Results showed that uranium adsorption peaked at 120 minutes, with a removal efficiency of 73.81% at 25°C and 10 ppm. Adsorption decreased with increasing U concentration but improved at elevated temperatures. Optimal pH for maximum adsorption was found to be pH 6 for 100 ppm and pH 4 for 50 ppm solutions.

In competitive adsorption experiments, the modified MSNs demonstrated a selective affinity for uranium over other heavy metals (Cr, Ni, Cu, Zn, Cd, Pb), with up to 40% uranium removal efficiency. These findings highlight the potential of polyphosphate-modified MSNs for selective and efficient uranium decontamination from contaminated water sources.

Keywords: Uranium removal, Mesoporous silica nanoparticles, Adsorption, Polyphosphate, Water treatment

The revitalization of the Radioisotope and Radiopharmaceutical Technology Facility (ITRR) is a critical step toward strengthening the domestic production of radioisotopes in Indonesia. Previously operated by the Indonesian Nuclear Industry (INUKI), the facility has been transferred to the National Research and Innovation Agency (BRIN) for rehabilitation following years of inactivity and contamination. This study implements a structured risk management framework based on ISO 31000:2018 and SNI 8615:2018 to identify and assess risks associated with the revitalization process. Radiological impacts are assessed based on IAEA Safety Report Series No. 77, while non-radiological risks are evaluated using national standard SB 006-1-BATAN:2019. Fault Tree Analysis (FTA) identifies eight primary risk causes, while the Hazard and Operability Study (HAZOP) reveals 18 detailed risks across operational domains. Each identified risk is addressed with a corresponding safety measure to mitigate its impact.
To support decision-making for cost planning, Monte Carlo Simulation is applied to estimate contingency costs related to decontamination, dismantling, and supporting facility maintenance. Using Oracle Crystal Ball, the simulation produces a mean contingency cost of IDR 21.7 billion, with a 90th percentile maximum of IDR 23.6 billion. These figures correspond to 5.04%–14.10% of the initial revitalization budget, providing a robust financial buffer against identified risks. This integrated approach demonstrates the value of combining qualitative and quantitative risk assessment tools to enhance project safety and financial predictability. The methodology and findings offer a replicable model for similar nuclear facility revitalization projects and reinforce the importance of proactive risk management in achieving sustainable outcomes in the nuclear energy sector.

Nuclear safety analysis, particularly in the context of severe accidents (SAs), requires modeling complex physical and stochastic phenomena. The Risk-Oriented Accident Analysis Methodology (ROAAM+) framework, under development at KTH, addresses this by integrating full-scale mechanistic models with computationally efficient surrogate models (SMs). However, many existing SMs lack the capacity to capture the inherently dynamic and time-dependent nature of severe accident scenarios. To address this, this study explores the use of machine learning (ML) techniques, focusing on neural operators, specifically Deep Operator Network (DeepONet), to develop SMs that can emulate the behavior of the MELCOR code – a comprehensive tool used for severe accident simulations.

This research addresses the use of several adsorbents in mitigating soil contamination of heavy metals and radionuclides. Many radiological or heavy metal sources emit harmful substances such as Lead, and Cesium. These dangerous substances accumulate in the soil, and plants absorb them through their roots. The objective of this research is to find a suitable combination of adsorbents and pollutants, in ideal conditions, and application methods. This was done through several quantitative analysis methods such as ICP-MS and SEM. These were utilized to measure the concentrations of the adsorbents and pollutants within the soil in several conditions. In addition to investigating the physical changes in the soil before and after remediation. Many samples were tested with varying soil depths such as 0-15cm and 15-30cm, and amounts ranging from 0.1-1g. Initially, the soil was tested alone with contaminants like Lead and Cesium to observe the highest retention rates. It was concluded that Lead was retained better by the soil. Further on, it was investigated which soil types worked best with which adsorbents; such as zeolite, bentonite, and silicate– in tackling the pollutants present. Once an ideal combination was found, such as Lead and silicate with 100% adsorption; a control variable was set with a variation in amounts of soil, pollutant, and adsorbent. This was created to test the adsorption isotherms and understand fully the model behind the adsorption that took place. The results show a promising future as the ideal conditions were found, which will move on to future work focusing on ideal application methods. On the other hand, Cesium has shown distinct results from its other samples with zeolite, reaching 86%, which can be explored to find optimum conditions.

Soil plays an essential role in the earth's surface systems, being an essential source of human radiation exposure and facilitating radionuclides' mobilisation into the environment. Investigating radioactivity levels in cultivation areas is necessary to safeguard soil, water, and air quality and foster sustainable development. The levels of natural radioactivity in the planting soils of medicinal plants remain scarce. Therefore, comprehensive data is needed to identify appropriate actions toward achieving the United Nations Sustainable Development Goals (SDGs) and safe and quality medicinal plants for traditional medicine and drug synthesis. This study aims to assess the potential radiological risks posed to the residents of medicinal plant cultivation areas in Pahang and Selangor, Malaysia, by evaluating various radiological parameters.

The Bandung TRIGA 2000 Research Reactor, a TRIGA Mark II-type facility upgraded to 2 MW in 1996, is currently undergoing preparations for decommissioning in accordance with national regulatory requirements. One of the critical components in this process is biological shielding, which serves as a protective barrier against neutron and gamma radiation. Accurate estimation of neutron flux distribution within this shielding is essential for assessing activation levels, classifying radioactive waste, and determining appropriate dismantling strategies. This study utilizes OpenMC Monte Carlo simulations to estimate neutron flux distribution across the biological shielding by segmenting the structure into 0.25 m³ concrete cells aligned with planned dismantling units.
Simulation results across operational configurations 93 to 100 reveal that neutron flux peaks occur near tangential and piercing beam ports, where values exceed 2.3×10¹⁰ cm⁻²s⁻¹, particularly in configuration 93. These localized flux maxima are attributed to neutron streaming effects, confirming the need for geometry-specific analysis. In contrast, outer shielding cells consistently register neutron flux below 10⁶ cm⁻²s⁻¹, indicating minimal activation potential. A clear attenuation trend is observed in both spatial and temporal domains, as neutron flux decreases with distance from the core and across successive reactor configurations.
These findings highlight the significance of incorporating spatial partitioning and historical operational data into decommissioning assessments. The high-resolution flux data enable targeted dismantling efforts, optimized waste classification, and enhanced radiological safety. Furthermore, the results demonstrate the utility of advanced simulation tools for reactor-specific decommissioning planning, reinforcing the need for tailored radioprotection strategies in zones with elevated neutron activation.

High-Temperature Gas-Cooled Reactors (HTGRs) represent a promising technology for decarbonizing both electricity and industrial heat sectors due to their high-temperature output, efficiency, and inherent safety features. This paper presents the final design of the HTGR-POLA, a 30 MWth prismatic block-type reactor concept developed at the National Centre for Nuclear Research (NCBJ) in Poland [1,2]. The primary goal of this work was to establish an optimized core configuration that ensures a long operational cycle, efficient fuel utilization, and robust safety margins under a wide range of operational and accidental conditions. The final design was achieved through a comprehensive, multi-parameter analysis involving iterative adjustments to fuel enrichment, TRISO particle packing fractions, and the strategic placement of burnable poisons. The reactor is designed to operate with helium as a coolant at 4 MPa, with inlet and outlet temperatures of 325°C and 750°C, respectively.

The world currently employs nuclear energy to meet the growing energy demand and contributes 10% of global electricity output. Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs) are the main types of reactors used in the nuclear power industry. Together, they are known as Light Water Reactors (LWRs) and account for over 90% of the world’s reactors, with PWRs making up more than 75% of them.

Ongoing advancements in nuclear reactor technology have led researchers to develop the Water-Water Energetic Reactor (WWER) or VVER. The evolution of VVER-1000 and VVER-1200 is the latest reactor design, featuring a higher power capacity of approximately 1200 MWe (gross) and upgraded passive safety measures.

Safety and risk mitigation are central to nuclear plant design and include numerous security layers to handle malfunctions and safely shut down operations. Even with these precautions, there remains a small chance of core damage and containment failure, which could lead to radiation exposure. PCTRAN's latest version can handle both DBAs and BDBAs, including faults like LOCA, core melt, and station blackout, making it a strong tool for assessing emergency response scenarios.

This analysis investigated the plant's reaction to combined internal and external steam line breaks alongside a turbine trip. In the first scenario (C1), each steam line break was assumed to have a 25% failure rate, while the turbine trip was set at 50%. In the second case (C2), all malfunctions were modeled at a 50% failure rate. The simulation examined changes in plant variables, such as reactor pressure, temperature, core water level, and thermal output, and how emergency safety systems were engaged in controlling hazardous conditions.

A 2 page-extended abstract, titled "Seismic Response Calibration of a Scaled CPR1000 Prestressed Concrete Containment Vessel Using ABAQUS" has been attached to the submission.

The increasing generation of disused sealed radioactive sources (DSRS) in Indonesia has created an urgent need for sustainable and secure disposal solutions. Deep borehole disposal (DBD) in limestone formations, especially from abandoned wells, presents a promising approach due to their high structural stability and low porosity, which effectively limit radionuclide migration. This study investigates the optimization of neutron-based porosity measurement methods using ³He, BF₃, and boron-lined NaI detectors in abandoned wells, addressing the impacts of variables including salinity, wax deposition, and porosity.
Simulations performed using Particle and Heavy Ion Transport code System (PHITS) revealed that ³He detectors provide the highest sensitivity to porosity changes and remain effective across a broad range of salinity levels, followed by BF₃ and boron-lined NaI detectors. Sensitivity was observed to increase with porosity but decline significantly with salinity levels of 5% to 30%, attributed to neutron absorption by chlorine ions in saline water. Wax deposition, modelled as a 3 mm layer, also reduced detector sensitivity, with ³He detectors showing the least impact (0.36%-16.12%), making them the most suitable for complex borehole environments. Boron lining improved thermal neutron capture efficiency but introduced detection sensitivity fluctuations, particularly in BF₃ detectors. Validation through PHITS simulations demonstrated accurate gamma energy spectra alignment with IAEA standards and consistent near-to-far detector count ratios when compared to reference datasets.
These findings highlight the need for advanced detection technologies and simulations to improve reliability of porosity measurements in DBD applications. The integration of such systems addresses critical challenges posed by salinity, wax deposition, and neutron interactions, supporting the efficient and secure management of radioactive waste. This research provides a significant contribution to the practical implementation of DBD, enhancing its viability as a sustainable disposal solution for radioactive waste in Indonesia and global.

To support the development of a Near-Surface Disposal (NSD) facility in Serpong Indonesia for the interim storage of low- and intermediate-level radioactive waste (LILW) generated by research reactors, nuclear industrial and medical activities in Indonesia, it is essential to study and forecast groundwater behavior for the safe disposal of radioactive waste to prevent contamination and guarantee long-term environmental safety. A NSD site is selected for low- and intermediate-level radioactive waste from research reactors due to its cost-effectiveness, engineered barrier feasibility, ease of monitoring, and suitability for shorter-lived waste. The pioneer assessment of technological NSD facility design criteria to meet the International Atomic Energy Agency (IAEA) standard was conducted, including groundwater parameters, such as depth, pH, and total dissolved solids (TDS). Notably, no comparable groundwater assessment for NSD systems using artificial intelligence currently exists. Therefore, in this study, effective monitoring and control require forecasting by using a robust time series forecasting model based on historical records of groundwater quality from 2012 to 2019. The study forecasted groundwater quality over 50 years. The predictive capacities of three machine learning models, such as Long Short-Term Memory (LSTM), Prophet, and Exponential Smoothing State Space (ESSS), were tested. Results showed that the Prophet model performed better with lower error metrics, including depth parameter at Mean Absolute Error (MAE)=0.71, Mean Squared Error (MSE)=0.77, Root Mean Squared Error (RMSE)=0.88, and Mean Absolute Percentage Error (MAPE)=7.41%, pH parameter at MAE=0.21, MSE=0.10, RMSE=0.32, MAPE=4.89% and TDS parameter at MAE=12.16, MSE=830.9, RMSE=28.82, and MAPE=31.62%. The study concluded that groundwater depth increases, pH decreases with time, and TDS changes seasonally, signifying the requirement for effective groundwater management strategies. The findings highlight the efficacy of machine learning models in supporting sustainable groundwater monitoring NSD sites over 50 years that can be beneficial for the long-term sustainable design of radioactive waste disposal.

The rapid expansion of nuclear energy has intensified the need for effective and secure methods of disposing of high-level radioactive waste (HLW). Deep geological repositories (DGRs) and deep borehole disposal (DBD) are two leading strategies for isolating high-level waste (HLW) from the biosphere. Despite their technical potential, both approaches face unresolved issues, including geological uncertainties, degradation of engineered barriers, and societal resistance. Underground research facilities (URFs) play a critical role in addressing these challenges by enabling in situ evaluation of host rock properties, engineered barriers, and radionuclide migration under repository-relevant conditions.
Global efforts in DGR and URF development have produced notable facilities such as Onkalo (Finland), Cigéo (France), and Beishan (China), each contributing to the advancement of thermal-hydraulic-mechanical-chemical (THMC) models and safety validation. However, technical limitations persist, including microcracking in host rocks, copper corrosion, and bentonite alteration, exacerbated by complex long-term environmental conditions. Non-technical barriers such as regulatory delays, political shifts, and community mistrust further complicate repository implementation.
Recent innovations focus on integrating digital twin systems, autonomous monitoring, and real-time stress sensors to enhance predictive modelling and safety assurance. Repurposing abandoned infrastructure, such as oil wells, has also emerged as a cost-effective and geologically feasible alternative for URF deployment. Concurrently, stakeholder engagement models, such as consent-based siting and community-driven monitoring, offer pathways to increase public acceptance and trust.
URFs remain indispensable to the future of HLW management, providing critical data and validation frameworks that are essential for effective management. The convergence of scientific innovation, regulatory clarity, and societal participation is crucial to achieving sustainable and ethically grounded solutions for nuclear waste disposal.

Radioactive waste from industrial activities, hospitals, research and decommissioning of nuclear/radiation facilities needs to be managed completely until it is stored and disposed of so that it is safe, sustainable and does not pollute the environment. In line with these waste management principles, Indonesia will prepare a long-term storage and disposal, as part of the back end of the radioactive waste management stage, which aims to isolate waste so that there is no radiation exposure does not exceed the dangerous threshold to humans and the environment. The required level of isolation can be obtained by implementing various storage and disposal methods, including the long-term storage (LTS), near-surface disposal (NSD) and borehole disposal (BHD). Research on the LTS site, NSD and BHD (integrated disposal) of radioactive waste is carried out to determine the selected potential areas and optimize them to meet safety criteria. For the initial stage, a study was carried out in the Jawa Island. The steps taken include: 1) determining site criteria based on legislation, IAEA recommendations and expert opinions, 2) acquisition and evaluation of secondary data, 3) field surveys, 4) laboratory analysis, 5) data analysis and evaluation of geology and environment for the LTS, NSD and BHD sites. Based on the evaluation results, several potential areas for LTS, NSD and BHD facility sites were found. The potential areas sequentially from west to east of Java Island include: 1) Serang, Banten (andesite rock), 2) Serpong Nuclear Area, Banten (claystone of Bojongmanik Formation), 3) Subang, West Jawa (claystone of Subang Formation), 4) Sumedang, West Jawa (claystone of the Subang Formation), 5) Rembang, Central Jawa (sandstone of the Wonocolo Formation) and 6) Tuban, East Jawa (claystone of the Kujung Formation and siltstone of the Tuban Formation).

Nowadays, charged particle accelerators play an important role as powerful engines that speed up tiny charged particles to incredibly high speeds close to the speed of light. These machines are essential in many scientific areas: Understanding the Universe, many medical applications, nuclear Studies, Astrophysics, and Industrial Uses. This research highlights on the longitudinal electron beam dynamics in a compact S-band standing wave radio-frequency photoinjector of 1.5 cell, focusing on the generation and acceleration characteristics of the electron beam under the RF fields. This research clearly examines the intricate interplay between electric fields and their profound impact on the extraction and acceleration mechanisms governing electron beams within accelerator cavities. Through meticulous analysis,
we are able to know how variations in electric field strength, under various conditions within the photoinjector setup such as the RF phase, directly affect the efficiency of beam generation, extraction and the subsequent acceleration processes within these specialized cavities. This is carried out with a support of analytical models and simulation program (MATLAB).
This study is also unique in the country and provides scientific access and resources for accelerator physicists, experts and researchers.

Natural flake graphite (NFG) is a cost-effective, carbon-rich material known for its high thermal stability and unique defect structures, offering considerable promise for application as a thermoluminescent dosimeter in the detection of ionizing radiation. This investigation focuses on the development of a thermoluminescence (TL) material for dosimetry that may improve the performance of current passive dosimeters. The study aims to analyze the essential TL characteristics of carbonaceous NFG. The dosimetric characteristics of commercially available NFG in response to X-ray photon-beams from Linac, at doses ranging between 2 Gy to 20 Gy, have been thoroughly investigated. The characteristics include TL glow curve, dose-response, sensitivity, energy dependency and fading. The results demonstrate that the NFG has a high linear response in the dose range under investigation and a higher sensitivity at lower doses. The NFG sample demonstrated remarkable reproducibility, with a standard variation of less than 3%. Fading study was performed under laboratory light and dark condition, revealed a minimum rate of fading for both condition. The SEM/EDX analysis confirms that NFG degrades microstructurally in a dose-dependent manner. X-ray diffraction (XRD) and Raman spectroscopy are used to evaluate the structural changes brought on by radiation doses. All of these investigations support the structural changes caused by photon irradiation. In conclusion, NFG exhibits great promise as a useful material for radiation dosimetry applications.

Keywords: Thermoluminescence (TL) Measurement, Raman Spectroscopy, X-ray Diffraction (XRD)

At the King Abdulaziz City for Science and Technology (KACST) in collaboration with the university of Hail (UOH), a beam line injector is being constructed to provide the multi-purpose low-energy applications. The injector is being equipped with a 90◦ high resolution mass analyzing selector magnet system and a new ECR ion source. The magnet system was designed to provide a singly-charged ion beam of kinetic energy up to 50 keV and ion mass up to 1500 amu with the mass resolution of ∆ m/m = 1/1500. In this paper, the ion-optical calculations, the degermation of the required momentum resolution and the actual analyzing system parameters will be discussed. The simulation of the beam envelope along the injector and through the magnet will be presented.

This paper focuses on thermal-hydraulic analyses to investigate safety-related aspects of the Reference LFR system, a conceptual Small Modular Reactor cooled by molten lead, using the RELAP5/Mod3.3 code in its version modified by the University of Pisa to account for heavy liquid metals as working fluid. The primary aim of this work is to develop a physically accurate model of the Reference LFR, capable of capturing both the thermal-hydraulic and neutronic behaviour of the reactor. This model is intended to be applied to simulate transient scenarios relevant for LFR operation, such as the Unprotected Loss of Flow (ULOF), Unprotected Loss Of Heat Sink (ULOHS), and Unprotected Transient Over-Power (UTOP).

Hideo Nakamura
Japan Atomic Energy Agency (JAEA)
Tokai, Japan
nakamura.hideo@jaea.go.jp

Ahmed Bentaib
Autorité de sûreté nucléaire et de radioprotection (ASNR)
Fontenay Aux Roses, France
ahmed.bentaib@asnr.fr

Martina Adorni
OECD Nuclear Energy Agency (NEA)
46, quai Alphonse Le Gallo, 92100 Boulogne-Billancourt, France
martina.adorni@oecd-nea.org

Abstract
The Working Group on the Analysis and Management of Accidents (WGAMA) of the OECD Nuclear Energy Agency (NEA) addresses safety issues of existing nuclear reactors and related technologies as well as emerging challenges on evolutionary and innovative reactor designs and nuclear technologies, including Small Modular Reactors (SMRs), related to their potential design-basis accident (DBA) and beyond design-basis accident (BDBA). The WGAMA coordinates research activities including workshops and technical publications in the fields of thermal-hydraulics (T/H), computational fluid dynamics (CFD) and severe accidents (SAs). Obtained knowledge is shared among member countries to improve confidence in the safety analyses that use computer codes required to represent accident phenomena that may arise during an accident in nuclear power plant (NPP) as precisely as possible.

This paper reviews the activities on the 3 main pillars: thermal hydraulics, CFD and severe accidents.
Acknowledgments
The Nuclear Energy Agency (NEA) Committee on the Safety of Nuclear Installations (CSNI) and the Working Group on the Analysis and Management of Accidents (WGAMA), wish to acknowledge the significant contributions of those individuals who had a key role and leadership in the conduct and success of the activities described in this report, especially for bureau members including past chairs, task leaders, contributors and members, and secretaries for their selfless contributions.